ML17055D923
| ML17055D923 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/02/1988 |
| From: | Jerrica Johnson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17055D922 | List: |
| References | |
| 50-220-88-15, 50-410-88-15, NUDOCS 8806150466 | |
| Download: ML17055D923 (26) | |
See also: IR 05000220/1988015
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report
No.
Docket No.
88-15/88-15
50-220/50-410
License
No.
DPR-63/NPF-69
Licensee:
Niagara
Mohawk Power Corporation
301 Plainfield Road
Syracuse,
13212
Faci l ity:
Nine Mile Point, Units
1 and
2
Location:
Scriba,
Dates:
May 6,
1988 through
May 24,
1988
Inspectors:
Approved by:
ate
W.A. Cook, Senior Resident
Inspector
W.L. Schmidt,
Resident
Inspector
A.G. Krasopoulos,
Resident
Inspector (Acting)
~/i,
f z
>:frp+
p'>>
.R. Johnson,
Chic
, Reactor
Projects
Section
2C,
INSPECTION
SUMMARY
Areas
~Ins ected:
Routine inspection
by the resident
inspectors
of station
activities including Unit
1 and
2 power operations,
licensee
action
on
previously identified items, plant tours,
safety
system walkdowns,
surveillance testing reviews,
maintenance
and modifications review, review of
Erosion-Corrosion
Program,
LER and Special
Reports
review and review of Unit
1
Appendix
R compliance.
This inspection
involved 137 hours0.00159 days <br />0.0381 hours <br />2.265212e-4 weeks <br />5.21285e-5 months <br />
by the inspectors
which included
15 hour1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> s of backshift inspection
coverage
and
7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> s of
weekend inspection
coverage.
P
Results:
Inspector review of recent fire barrier penetration
concerns identified
an apparent
VIOLATION of 10 CFR 50 Appendix
R.
Inspector
review of this issue
is documented
in Section
10.
The licensee's
quality control measures
for the
installation of fire penetration
seals
was not available for inspector .review
prior to the conclusion of the inspection
and this item will remain
UNRESOLVED
ending review of the documentation
the licensee
has committed to provide the
nspectors
(also
see Section
10).
A review of the Unit
1 Erosion-Corrosion
rogram is documented
in Section
9.
8806150466
880607
ADOCK 05000220
O
DETAILS
Review of Plant Events
(71707,93702,71710)
UNIT
1
During this inspection period,
the unit remained
shutdown for the
1988
refueling outage.
The core
remained off-loaded pending final resolution
of numerous
Inservice Inspection
concerns identified prior to and during
the outage.
On May 10,
the licensee
determine'd that the Core Spray System
high p'oint
vent valve
( IV 40-30) did not stroke closed
in less
than the
30 seconds
Technical Specification
(TS) limit.
IV 40-30 is
a normally closed,
inside containment isolation valve.
Actual stroke time was measured
to
be 32.28
seconds.
This was discovered
as
a result of licensee
followup
to
Inaccurate
Closed Position Indication
on Motor-Operated
Valves.
IV 40-30 was the only valve found to have
a stroke time in excess
of the
TS limit.
The inspector determined
from the licensee that the most
probable
cause of this excessive
closing time is improper placement of the
position indication limit switches.
The inspector will review licensee
actions to correct this problem in a subsequent
inspection.
On May 12,
the licensee
preliminari ly determined that the reactor coolant
system pressure
versus
temperature
curves in the
TS may be inaccurate.
Information provided to the
NRC staff by the licensee,
to date,
indicates
that the reactor vessel
irradiated
samples
may not be of the
same material
as the vessel beltline
~
The irradiated
samples
are
used to estimate
the
vessel
metal radiation embrittlement.
This information is then
used to
assist
in generating
revised reactor
vessel
heatup
and cooldown limits.
The licensee
is evaluating this concern
along with their engineering
consultant
MPR and General
Electric Co.
Evaluations
conducted,
to date,
indicate that the current
TS limits and operating
curves
are conservative.
The inspectors will continue to monitor licensee
progress
on this issue.
On May 12, the licensee
concluded that there is a significant discrepancy
between
the
FSAR and another
licensee classification,
namely,
a
1983
"Appendix
B Determination" concerning whether the emergency battery
motor-generator
set battery chargers
(161 and
171) are safety or
non-safety related.
The
FSAR considers
the battery chargers
to be vital
(safety related) to the restoration of emergency lighting and
DC power
from the emergency
diesel
generators
within 130 seconds
following the
Contrary to this, the
1983 "Appendix
B
Determination" concluded that the battery chargers
were non-safety
related
because
the batteries
have sufficient capacity to supply
DC loads following a
LOP.
The licensee
has considered
the
161 and
171 battery chargers
non-safety
related
since
1983
and all maintenance
and
spare
parts were treated
accordingly.
As of the
end of this inspection period,
the licensee
had
not resolved this concern.
The .inspectors will review this item in a
subsequent
inspection period.
1.2
UNIT 2
During this inspection period the unit was in
a planned
maintenance
outage
which began
on April 29.
The outage
was concluded
and the reactor
was placed in the
STARTUP Mode on
May 22,
and taken critical at 11:46
p.m.
on the
same
day.
a.
On May 10, the
Emergency
Response
Facility (ERF) computer
was taken out
of service for planned modifications.
The licensee notified the
Headquarters
Duty Officer of this
eve~,
via the
ENS because
the licensee
considers
the
ERF computer function essential
to emergency
response
assessment.
By licensee
standards,
the
ERF computer out of service
constitutes
a major loss of emergency
assessment
capability.
A similar
notification was
made
on
May 24 when the Liquid Radwaste
Processing
computer
was out of service
due to
a hardware
problem.
The
ERF computer
function is part of the larger Liquid Radwaste
Processing
computer
system.
b.
The inspector discussed
the reportability of this event with licensee
management.
The licensee
stated that the practice of reporting the
computer
being out of service is
a carryover from Unit 1.-
Because
of the
redundancy
in emergency
assessment
capability available at Unit 2 via the
process
computer
and telephone
communications,
the licensee
is
reconsidering
the reportabi lity of the
ERF computer not being available
for service.
The inspector will review the licensee's
final decision.
On May 18, four (10') emergency
response
were out of service
and
the licensee notified the
NRC via the
ENS of the degraded
emergency
response
notification system condition.
The licensee
had planned
an
outage of three sirens
due to the necessity
for work on Niagara
Mohawk
transmission
lines.
However, the drawings that the licensee
referred to
were inaccurate
and
a fourth siren
was deenergized
along with the three
planned.
The inspector determined
that the licensee
is revi-sing and verifying all
emergency notification system electrical
drawings to prevent recurrence.
The inspector will review licensee corrective actions
when completed.
The inspectors
reviewed licensee
preparations
for restart of Unit 2 and
verified selected
prerequisites.
No discrepancies
were noted,
d.
On May 23, with the reactor at approximately three percent
power and 400
psig, the
A recirculation
pump outer seal
pressure
went to zero (0) psig
and the high pump seal
leakage
alarm sounded.
A reactor
shutdown
was
commenced at 9:50 a.m.
and the
A recirculation
pump was secured
and
isolated
by 10:02 a.m.
At 10: 17 a.m.
the licensee
declared
an
UNUSUAL
EVENT following an increase
in drywell floor drain leakage
in excess
of
the
TS limit of five (5)
gpm.
Leakage
was below the
TS limit
approximately
one half hour later
and the licensee
secured
from the
UNUSUAL EVENT.
The reactor
was placed in HOT
SHUTDOWN by 2:34 p.m.
The inspectors
reviewed licensee
actions in response
to this event
and
'ound
no discrepancies.
The inspectors will review the licensee's
evaluation of the seal failure in
a subsequent
inspection period.
The inspectors verified that the licensee
made the appropriate
10 CFR 50.72 notifications via the
ENS for the events
discussed
above.
2.1
~Foliowu
on Previous Identified Items (71707,82203)
Unit
1
(Closed)
Inspector
Followup Item (50-220/83-04-01):
Complete installation
of two additional
emergency notification sirens
as part of the system
enhancement
program.
The inspector verified that the licensee installed
two additional sirens,
one in Oswego
and one South of Ninetto,
as
required.
This item is closed.
Plant
~ins ection Tours (71707,71710,62703,64704,71881)
During this reporting period,
the inspectors
made tours of the Unit
1 and
2 control
rooms
and accessible
plant areas
to monitor station activities
and to make
an independent
assessment
of equipment status,
radiological
conditions,
safety
and adherence
to regulatory requirements.
The
following were observed:
Unit
1
No discrepancies
were
noted'.2
Unit 2
While conducting
a tour of the Reactor Building, the inspector identified
a few minor housekeeping
items which were brought to the attention of the
control
room operators.
The inspector
subsequently
verified that these
items were corrected.
The inspector also noted the imprope'r installation
of SILTEHP protective wrap around the
3A squib valve electrical
cable.
This item was brought to the attention of the Fire Protection Supervisor
who took action to correct this discrepancy.
The protective wrap is
installed
on one squib valve cable
because
of the requirement for
divisional separation.
The Fire Protection
Supervisor also initiated a
maintenance
procedure
revision to ensure
the protective wrap is properly
restored after
squib valve maintenance.
No violations were identified.
Surveillance
Review (61726)
The inspectors
observed
portions of the surveillance
testing listed below,
to verify that the test instrumentation
was properly calibrated,
approved
procedures
were used,
the work was performed
by qualified personnel,
limiting conditions for operations
were met,
and the system
was correctly
restored
following the testing.
4.1
Unit 1
Nl-ISP-Q-68, Reactor Building-to-Torus Vacuum Relief Valve
Instrumentation
Testing,
performed
on May 12,
1988.
During this inspection
period the licensee
conducted
the monthly
operability surveillance
on the
102 emergency diesel
generator.
During this surveillance test,
the engine tripped
on low lube oil
pressure.
The pressure
sensor is located
on the side of the crank-
case.
The licensee
has investigated
the problem
and believes
that the pressure
sensing
was impinged
upon
by lube oil
spraying
from the crankcase
internal
lube oil relief valve.
The
licensee
plans to inspect
the
103 diesel for the
same condition.
The inspectors will review licensee
findings in a subsequent
report.
4.2
Unit 2
N2-FSP-FPG-R002,
Halon System Nozzle Flow Test,
performed
on May 13,
1988.
No violations were identified.
5.
Maintenance
and Modifications Review (62703,37700,37701)
The inspector
observed
portions of various safety-related
maintenance
and
modification activities to determine that redundant
components
were
that these activities did not violate the limiting conditions
for operation,
that required administrative
approvals
and tagouts
were
obtained prior to initiating the work, that approved
procedures
were
used
or the activity was within the "skills of the trade", that appropriate
radiological controls were implemented,
that ignition/fire prevention
controls were properly implemented,
and that equipment
was properly
tested prior to returning it to service.
5.1
Unit
1
The inspectors
reviewed various aspects
of the
Emergency Battery ll and
12 replacement.
The inspector s observed that this modification was well
planned
and executed.
No discrepancies
were noted.
5.2
Unit 2
Prior to restart of the unit on May 22, the inspectors. reviewed licensee
startup preparations
and prerequisites.
One item reviewed
by the
inspectors
was the Operations'taff modification notebook.
This
notebook
was required reading for the entire Operations staff and
included
a
summary of each modification completed during the planned
outage.
The inspector
determined that the notebook adequately
summarized
the significant outage modifications,
provided appropriate modification
5.3
details
where necessary,
and was reviewed by shift personnel
prior to
unit startup.
The inspector
found no discrepancies.
Review of Maintenance
Self-Assessment
- Unit
1 (90713)
On May 20, the inspector held
a discussion with licensee
representatives
to review their Nine Mile Point Unit
1 Maintenance
Self-Assessment
conducted
in 1987.
This self-assessment
was
an Institute of Nuclear
Power Operations
( INPO) initiative designed
to accelerate
maintenance
performance
improvements
in the nuclear industry.
The self-assessment
conducted
by Niagara
Mohawk personnel
is the first phase of the assessment.
The second
phase is
a special on-site maintenance
review and assessment
by
INPO representatives
(the
second
phase
has not been
conducted,
to date).
A general
overview of the self-assessment
was presented
to the i nspector.
The licensee
indicated
the assessment
was useful
in identifying both
specific
and program weaknesses.
The identified problems will be tracked
by the licensee
and corrective action taken
as appropriate.
The licensee
stated
many
new initiatives,
such
as the materials
issue
and control
program enhancements,
have already begun'pecific
self-assessment
findings were not discussed
at this meeting.
The inspector will follow
selected
items
and discuss their progress with the licensee
in a
subsequent
irspection period.
The inspector
had
no further questions.
~Ph sical ~Securit
Review (71709)
The inspector
made observations
to verify that selected
aspects
of the
station physical security program were in accordance
with regulatory
requirements,
physical security plan
and approved
procedures.
The inspector walked
down the perimeter
fence to verify that there
were
no obstructions
in the vicinity of the fence or other fence
impairments
that could aid the unauthorized
entry of an individual into the plant.
No unacceptable
conditions were
identified'eview
of Licensee
Event
~Re orts ~LERs
and ~Secial
~Re orts
~SRs
(90712,92700),
These
LERs and
SRs submitted to the
NRC were reviewed to determine
whether the details
were clearly reported,
the cause(s)
properly
identified and the corrective actions appropriate.
The inspectors
also
determined
whether the assessment
of potential
safety
consequences
had
been properly evaluated,
whether generic implications were indicated,
whether the event warranted
on site follow-up, whether the reporting
requirements
of 10 CFR 50.72 were applicable,
and whether the
requirements
of 10 CFR 50.73
had been properly met.
(Note: the dates
indicated are the event dates)
0
7.1
Unit
1
a.
The following reports
were reviewed
and found to be satisfactory:.
LER 88-10,
4/19/88 - Failure to submit
a Special
Report within
thirty days.
SR, 5/6/88 - Fire'etection
and suppression
systems
for greater
than fourteen days.
b.
For the following report,
the licensee
has committed to issue
a
supplemental
report.
This report will be reviewed in a subsequent
inspection period:
LER 88-12,
4/18/88 - Failure to hydrostatically test
a portion
of the
ASME Class
1 Pressure
Boundary due to procedural
error.
7.2
Unit 2
a.
The following LERs were reviewed
and found to be satisfactory:
LER 87-74,
Rev.
1, 12/19/87
Inoperable fire barrier due to
breached floor plug.
LER 88-20, 4/7/88,
Secondary
Containment isolation
and
Standby
Gas Treatment
automatic start
due to spiking on normal
Reactor Building ventilation radiation monitor .
8.
~Re ort Unit
1 (90712)
On Oecember
19,
1987, Unit 1 was operating at
98% power when vibration in
the Feedwater
System
(FWS) piping resulted in the control
room operators
scramming
the reactor.
Subsequent
investigation of this event identified
the
FWS piping vibrations to be the result of flow oscillation caused
by
the
stem
and plug separation
of the
13A feedwater control valve.
Analysis of the
13A control valve failure, resultant
FWS transient
and
FWS piping and support
damages
are
summarized
in
a licensee
report to the
NRC dated
March 1,
1988 (NMP1L0229).
The inspector
reviewed the contents
of the March 1,
1988 report and found
no significant discrepancies
from the information independently
gathered
and assessed
by the
NRC staff.
The inspector
found the report to be
well-organized,
thorough
and concise.
Licensee. corrective
actions
outlined in the report appear to be appropriate.
The inspector specifically reviewed the following report attributes:
Scope of
FWS piping inspections,
including drywell piping
examinations.
Visual inspection
adequacy
and examination results.
Loose parts safety analysis
(pump impeller blade piece).
Sequence
of events,
including shaft driven feedwater
pump
impeller'ailure.
FWS repairs
and prerequisites
for unit restart.
Plant
systems
response
to
FWS transient,
including fire detection
system
response.
Correlation of this
FWS transient/failure to previous
FWS problems.
Operating/maintenance
history of feedwater control valves.
Analysis of No.
11 feedwater
pump suction piping corrosion.
Examination results of No.
12 feedwater
pump suction piping.
Results of feedwater control valves
No
~
11 and
12 internals
inspection.
Metallurgical analysis of stem-to-di sc failure.
FWS transient analysis results.
Although not specifically addressed
in the licensee's
corrective action
for this transient,
the inspector determined that in addition to the
improved stem-to-disc
weld design,
the licensee
is evaluating
new
feedwater control valve designs.
New designs
are being entertained
to
address
the valve plug and
cage
wear concerns
which contributed to the
valve failure.
Additional inspector observations
and findings have
been
documented
in
Region I Inspection
Report 50-220/88-02.
The inspector
had
no further
questions.
Review of Erosion-Corrosion
~Pro
ram
Unit
1 (61726)
On May 20, the inspectors
met with licensee
Engineering representatives
to discuss
the Carbon Steel
and
Low Alloy Piping System Erosion-Corrosion
Review Program for Unit
1 and its implementation during the current
1988
Refueling Outage.
The purpose of the program is to evaluate
the
performance of carbon steel
and low alloy piping systems
which are not
reviewed per the Inservice Inspection
Program,
but which are susceptible
to deterioration
caused
by high-energy single
and two-phase fluid
erosion-corrosion.
The program
was developed
as
a result of recent
industry problems,
NUMiARC initiatives and interest
by the
NRC staff in
programs established
by the nuclear utilities to address
this potential
high energy piping system failure mechanism.
(Reference,NRC Bulletin No.
87-01).
The inspector
determined that the Unit
1 program was formally implemented
for the first time during the current refueling outage.
Some feedwater
system piping wall thickness
data
was taken during earlier refueling
outages;
however, this information was not formally evaluated
and trended
in the fashion the
new program
has established.
The licensee
has
identified specific examination points in both single-phase
and two-phase
flow systems (ie. feedwater,
feedwater heating,
extraction
steam,
steam
drains, etc.) which will be baselined
during the
1988 outage
and
periodically examined
in the future.
Piping examinations
are currently
being performed
independent
of the ISI staff efforts by contractors with
direct oversight
by Niagara
Mohawk engineers.
-9"
The inspector
found the program to be well-structured
and adequately
detailed to ensure
consistent
examination results.
The inspector
questioned
the licensee
representatives
on the mechanisms
used to elevate
identified problems or potential piping degradation
to station
management
for resolution prior to unit restart.
The licensee
indicated that the
Erosion-Corrosion
Program does
have provisions for interim status
reports
to identify any significant potential
problems to management,
but no
specific guidance is provided in the program to document
and track
nonconformances
identified during the field examinations
or the
Engineering
review process.
This observation
was discussed
with station
management
who indicate that the guidance to track Erosion-Corrosion
Program identified nonconformances
would be strengthened
to ensure
overall
program results
are properly assessed
by station
management prior
to unit restart.
The inspector
had
no further questions.
10.
Review
Unit 1 (64704)
On March 26,
1988, while performing
a modification to replace
DC cables
from the Battery Board
Rooms
11 and
12, the licensee
determined that
some
of the existing fire barriers
were inoperable.
The inoperable fire
barriers (floors of the battery
rooms) were found to contain penetrations
sealed with unqualified material.
The licensee
documented this event
Licensee
Event Report
( LER) 88-09 and committed to review all Technical
Specification
(TS) fire barriers for adequacy
and repai r all non-functional
identified prior to restart.
The repair work and fire barrier
review began prior to the issuance
of the
LER.
In addition,
compensatory
fire watches,
as specified
in the TS, were established
until all fire
barriers
were reexamined.
The licensee's
evaluation of the inoperable barriers
concluded that this
problem was caused
by an inadequate
original (1983) review of the barriers
by the contractor hired to survey the barrier penetrations.
This contractor,
because
of inadequate
guidance
from Niagara
Mohawk, failed to identify and
list all of the penetrations
that were in the barriers.
Additionally,
this error was perpetuated
by an inadequate
surveillance
procedure.
The
surveillance
procedure directed the individuals performing the surveillance
to inspect the operability of a specific penetration
rather than the
operability of the barrier
as
a whole.
Thus,
the procedure
had
no provisions
to inspect the operability of the entire barrier,
only failures of penet-
rations listed in the procedure.
To prevent recurrence
the licensee is
updating the procedure
to include all the penetrations
that the original
survey failed to identify and is revising the surveillance
method
so that
the overall fire barrier adequacy
is inspected
rather than the operability
of a specific penetration.
To assure
that the penetration
surveillance list is accurate,
the
licensee,
as committed to in the
LER, is resurveying all of the
TS
barriers
and penetrations.
As of the end of this inspection period,
0
-10-
approximately
20% of the fire barrier reexaminations
have
been
completed.
Of the
1325 penetrations
inspected,
68 were found to be non-functional.
is considered
to be non-functional
by the licensee if its
seal
has
been
damaged
or if it has
been
sealed with materials
or methods
that have not been fire tested.
The fire tests
provide assurance
that
the installed
seal configuration will not degrade
during
a fire.
The inspectors
observed that the inoperable
were in fire
barriers that separate
safe
shutdown
components.
The barriers
in
particular are the Battery
Room floors, the Cable Spreading
Room floor
and the Auxiliary Control
Room floor.
A design basis fire in any of
these fire areas
has the potential of propagating
to the adjacent fire
areas
via the degraded
and could damage
the control
room and
both remote
shutdown
panels
(RSPs).
The
were installed to assure
the ability to safely
shutdown
the
plant in the event of a fire in the control
room.
Manning of the
is required
by Special
Operating
Procedure
for Control
Room Evacuation,
N1-SOP-9.
This procedure
requires that the operators trip the reactor
and initiate and control
emergency
cooling at the
RSP if that could not
be accomplished
in the Control
Room.
The rate of reactor
cooldown is
regulated
from these
panels
so that it does
not exceed
100~ F/hr.
The possibility that
a design basis fire has the potential of damaging
both
and the control
room is an apparent violation of 10 CFR 50,
Appendix
R,Section III.G.
Section III.G requires that fire protection
features,
such
as fire barriers,
be provided to protect
safe
shutdown
systems
and components.
These
features
must
be capable of limiting the
fire damage
so that at least
one safe
shutdown train remains
free of fire
damage.
APPARENT VIOLATION (50-220/88-15-01)
.
Detailed review of this concern
by the inspectors
determined that
a fire
which could damage
the Control
Room and both
RSPs is not
a highly
credible event.
This determination
was based
on several
mitigating
factors.
The areas
where the degraded
were'found were
protected with automatic
suppression
and detection
systems.
The
combustible
loading is composed
mostly of cable insulation with the
heavier concentrations
in the auxiliary control
room and cable
spreading
room.
Each of these
areas
is protected with two suppression
systems.
The auxiliary control
room is protected with a total flooding halon
suppression
system activated automatically
and
a total flooding C02
system that requires
manual activation.
The cable
spreading
room is
protected with an automatic
C02 total flooding system
and
a sprinkler
system.
The inspector verified these
systems
were operable.
Another
mitigating factor is that although the penetrations
have
been declared
they do contain seals
which provide
a degree of protection
and
resistance
to the spread of fires.
The licensee
has stated that even if this event occurred (ie.
a design
basis fire) the safe
shutdown
systems
would still work and provide
shutdown capability.
In
a design basis fire, one function that could be
lost is the condensate
return flow control
from the emergency
condensers
to the reactor.
This system regulates
the cooldown rate of the reactor.
Also, the capability to monitor the reactor pressure
and level
from .the
RSP could be lost.
However,
these
parameters
can
be monitored
from other
locations
in the plant.
In addition,
the starting
power source for the
emergency diesels
could
be lost.
However,
the diesels
are not relied
upon for the
HOT
SHUTDOWN phase
and the licensee
has procedures
and
materials,
in place,
to make the required repairs.
These
procedures
and
repair methods
were reviewed
by
NRC inspectors
during
a previous
inspection
and were found acceptable
(reference
NRC Region I Inspection
Report 50-220/85-01).
In addition to the degraded barriers,
the
new survey conducted
by the
licensee
has identified
a number of minor discrepancies
that are being
addressed
and corrected.
These discrepancies
include cosmetic
damage
to
the seals,
the presence
of duct seal material
on top of the fire seal
and
sealed
to
a depth slightly less
than that specified
in the
drawings,
but still adequately
sealed
to qualify as
a three
hour barrier.
The licensee,
as
a conservative
measure,
declared
these
and established
compensatory
measures
prior to dispositioning
the discrepancies.
The licensee's
disposition of these
concerns
involves
the review of the discrepancy
by
a Fire Protection
Engineer,
who makes
a
determination
of the adequacy
of the seal.
The
NRC inspectors
reviewed the evaluation
methods
used
by the licensee
to verify that
each identified discrepancy
is properly dispositioned.
During this review the inspectors
raised the concern that the licensee'
Engineering staff was declaring
untested
by
evaluations
performed after
a nonconforming condition was identified.
Unless
the evaluation
was in place at the time the non-conforming
condition is identified, the penetration
should
be declared
and reported to
NRC accordingly.
The inspectors
also requested
to review installation records of the
seals
to verify that the seals
were installed
as
per
the
design details.
This information was not available for review by the
inspectors
by the end of the inspection,
therefore this item will
remain unresolved
pending
a review of the records
being gathered
by the
licensee.
UNRESOLVED ITEM (50-220/88-15-02)
10.1
~Sommar
of ~Findin
a
The resurvey of the fire barriers,
to date,
has identified 68 non-functional
This survey also identified a number of minor
deficiencies
on the penetration
seals that require repair.
These findings
-12-
are
a cause for concern
because
the barriers
have
been repeatedly
(and
'rogrammatically)
inspected
and these
degradations
were not identified.
The degraded
barriers
represent
an apparent violation of the requirements
of 10 CFR 50, Appendix
R, which stipulate that fire protection features
such
as fire barriers shall
be provided to assure that at least
one shut-
down train remains free of fire damage.
At the conclusion of the inspection
the licensee
committed to provide the
NRC with assurance
that all penetration
installations
conform to the
design details.
Unit
1 Restart
Plan
~Meetin
(30702)
Following the
SALP Management
Meeting held at the Nuclear Training Center
on May 10,
1988,
the
NRC (Region I and
NRR) staff held
a meeting with the
licensee
(corporate
and station
management
personnel)
to discuss their
plans for restart of Unit l.
A tentative list of certain discussion
topics
for the meeting
was communicated
to the licensee
by
NRC letter dated
May
4,
1988.
These topics were reviewed
by the licensee
and
a status
was
provided during the meeting.
The licensee
plans to meet with the
NRC staff
three
weeks prior to Unit
1 startup to discuss
additional details of
management
controls of all items affecting unit startup.
Currently an
interim meeting is scheduled
for June
21,
1988.
~Meetin
With Local Of'icial (94600)
On May 13,
1988 the Nine Mile Point and FitzPatrick resident
inspectors
met with the Mayor of the City of Oswego.
The purpose of the meeting
was
to familiarize the Mayor with the role of the
NRC resident
inspectors.
Topics discussed
included resident
inspector
coverage of routine
and
off-normal plant events,
NRC and licensee
Emergency
Plans,
the
NRC
Inspection
Program
and the Systematic
Assessment
of Licensee
Performance
process.
No additional
concerns
were identified during this meeting.
Assurance
of ~ua'lit
Identification of. the Unit
system pressure,
versus
temperature
curve
and battery charger
problems represent
good detailed
engineering
reviews
by the licensee's
staff.
However, the errors which
led to these
concerns
appear
to be additional
examples of past
insufficient staff oversight of contractor activities.
The maintenance
self-assessment
appears
to be
a fruitful initiative and warrants
management
attention
and follow-through.
The Unit
followup efforts and
summary report are of good detail
and thorough.
The
erosion-corrosion
program appears
to be well-structured
and properly
implemented.
As noted in Section
10-; 1, the fire barrier penetration
problems
are of serious
concern
from a program implementation
standpoint.
Licensee
response
to these
concerns
appear
to be appropriate,
to date.
Exit ~Meetin
s (30703)
At periodic intervals
and at the conclusion of the inspection,
meetings
were held with senior station
management
to discuss
the
scope
and
findings of this inspection.
Based
on the
NRC Region I review of this
report and discussions
held with licensee
representatives, it was
-13-
determined that this report does not contain 'Safeguards
or 10 CFR 2.790
information.