ML16341F575
| ML16341F575 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 02/06/1990 |
| From: | Huey F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML16341F573 | List: |
| References | |
| 50-275-89-33, 50-323-89-33, GL-81-12, NUDOCS 9003050128 | |
| Download: ML16341F575 (26) | |
See also: IR 05000275/1989033
Text
U.S.
NUCLEAR REGULATORY COt1Y>ISSION
REGION
Y
Report
No. 50-275/89-33
and 50-323/89-33
License
Nos.
DPR-80 and
DPP,-82
Licensee:
Pacific
Gas
ard Electric Company
77 Peale Street
Room
1451
San Francisco, California 94106
Facility Name:
Diablo Canyon
Power Plant, Units
1 and
2
Inspectin('> at:
Diable Canyon
Power Plant, Units
1 and
2
InsII(ctior, Conducted:
December 4,
1989 to January
17,
1990
Inspecto> s:
P.
P. Narbut, Senior Resident
Inspector
t~.
P. Y>ilier, Reactor
Inspector
C.
8
,
In pector
A) roved b:
'f .
R. Huey,
C ief,
ngineering Section
(Jzo
IIa
e Si~ne<l
Summer:
Insperti
C
on During the Per.iod of December
4,
1989 through January
17,
1990
(R( port No, 50-F75/89-33
and 50-323/89-33)
Areas
Irsyecte(s:
The inspection
included follow-up of open items
and
a review
oOO>corisee act)on.
in response
to those
open items.
NRC Inspection
procedures
30703,
64704,
l'1-2515/87
and
92701
were
used during this inspection.
Results of Ir>~section ard Genieral
Conclusions:
1.
A violation was identified concerning
a lack of administrative controls
requiring operability of the positive displacement
charging
pump, which
is required for safe
shutdown in the event of a fire.
NRC guidance in
the
form> of Generic Letter 81-12 was available,
and would have helped the
licensee
to avoid this violation, had it b'een carefully reviewed
and
implemented.
2.
53 discrepancies
between'he
as-built plant and the approved fire
protection program were identified by the licensee,
some
as early as
1982.
The program should have
been
changed to resolve these
discrepancies
according to the license condition, which requires
evaluation
pursuant to
The licensee
stated that
NRC guidelines
in Generic Letter 86-10 were unclear,
and this
was given
as the reason for the lack of
resolutio() of this'issue.
However, the license condition takes
precedence
over Generic Letter 86-10.
iy(y(>Q()~i(.)lZS
~ o()02() /
pDO( (~
Q)~i()OO~
0
In conclusion,
the inspectors
found that the licensee failed to properly
implement the requirements
of the fire protection
program in two instances.
Summary of Violations or Deviations:
Two deficiencies identified during this inspection
appear to involve
violations of NRC requirements.
1.
Technical Specifications 6.8. 1 requires
procedural
implementation of
Regulatory
Guide
(RG) 1.33 items,
one of which is the fire protection
program.
The fire protection
program requires
the positive displacement
chargino
pump
(PDP) to be operable
so it can provide charging for safe
shutdown in the event,
a fire disables
both centrifugal charging
pumps.
Contrary to these
requirements,
there
were
no administrative controls
requiring the
PDP to. be operable to ensure
safe
shutdown.
This violation
is'ocumented ir the attached
n
>>
~
Technical Specifications
6.S.3
requires
procedural
implementation of
RG 1.33 items,
one of which is instrument calibration.
Licensee
procedure
AP C-450 "Preventative Vaintenance
Program" requires that
instrumentation
be entered
in the Recurring Task Scheduler
and calibrated
within the required interval.
Coptrary to these
requirements;
instruments for battery
11,
12, ]3, 21, 22,
and
23 voltage
and current
inc'icators
were past
the three year calibration iqterval.
Also, Unit
2,
~kV bus
F anc I-'oltage and Unit 2 battery 21, 22,
and
23 voltage
and
current were rot documented
in the Recurring Task Scheduler.
This
violation'is docL!o,ented
as
a non-cited violation in this inspection
report.
GIaei. Items
Su~mmat:
Thirteei: open items
v,ere reviewed.
Nine were closed,
four remain open.
e
DETAILS
Persons
Contacted
PGKE
- J
%f,'pl
- T
- )ts
- 8
- p
- D
g:T
<>> [~
- D
- E
a,J
~V,.
p
fi.
p..
R.
f>> ~
p.
R.
lt.
S
.A.
U ~
R.
ll.
p.
Townsend,
Plant Yanager
Angus, Assistant Plant Vlanager, Technical
Services
,
Kelly, Regulatory
Compliance
Engineering
Allen, NECS, Project Engineer
Crockett,
APY', Supervisor
Services
Rinkacs,
Regulatory Compliance
Roller, System Engineer
Taggarf., Director, Quality Support
RapI:, Ctiairman, Gnsite Safety
Review Group
Giffin, Assistanf, Plant Vanager,
Ylain Services
Ylikliesh, Assistant
Plant Hanager,
Operations
Services
Connell,
NECS, Project Engineer
Shoulders,
NKS,
GPEG Project
Engineer
C., Barkhuff, Acting
QC Ylanager
Yap, System
Engineei
SmitlI,
NECS Project Engineer
Vashington,
fIanager,
18C Ylaintenance
Johansen,
I80 Yaintenance
Baur, Supervisor,
Electrical Vaintenance
Panero,
NGS, *Fire Protection
Engineer
Kohnut, Supervisor,
Emergency/Safety
Services
Iyer, Design
Change
Engineer
Htmilton, Design
Change
Engineer
Nich'olson, Nuclear Safety
and Regulatory Affairs
Fuhriman, Quality Control
Spoutz,
System Engineer
Clark, Supervisor,
Ylechanical
Engineering
farrad-', Enigineering,
Yiechanical
Kao, Lead, Safety Pelated
Vlechanical
- *Attended the Exit Vieeting on December 8,
1989.
Licensee Action on Previously Identified Items
A.
0 en)
0 en Item 50-275/87-27-04,
Resolution of Differences
pr
p,pe
~a ~p p~~
in an ear1ier report,
an inspector
note
thattt
e licensee
identified several
(50) specific discrepancies
between
the
NRC
approved fire protection
program and the as-built plant
configuration.
Yiore significant examples
are:
30 ft rather thari
the approved
50 ft separation
between trains,
2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> penetration
seals
which reduce fire protection of the approved
3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire
barriers
in which they are installed,
and
some safe-shutdown
circuits without the approved
2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> fire rated barriers.
The
licensee
had documented
evaluations
of each
oi- these
discrepancies
ano consioered
each
%o be satisfactorily resolved
because
,/
compensating fire protection (detection
and supression)
is provided
and
because
the licensee
considers
the existing fire protection
adequate..
Based
on these evaluations,
the licensee
concludes
that
the level of fire protection in the plant is acceptable.
Since discrepancies
exist between
the approved
program
and the
plant, the licensee
should-change
the program in accordance
with the
applicable
license condition.
This condition requires that changes
to features of the approved
program
be evaluated
according to the
requirements
of 10 CFR 50.59;
and where these
changes
do not conform
to the criteria of 10 CFR 50.59,
the evaluation
should
be submitted
to the
HRC staff for review and approval.
Concerning
the
evaluations,
the inspectors
noted the following:
( I)
The evaluations
did not specifically address
the criteria of 10 CFR 50.59.
(2)
The licensee's
assumptions
concerning fire hazards,
loading, fire barriers,
and propagation of smoke
and hot gasses
did not appear to be consistent with the guidance of Branch
Technical
Position
(BTP)
CMEB 9.5. 1, which provides
Appendix
R guidance
concerning evaluation of fire protection.
In this regard,
the licensee
stated that transient
and in-situ
combustible
loading
was computed
using the licensee's
Fire
Protection
Database
System
(FPDS).
This is
a database
compiled
from plant walkdowns.
It also
computes
values for combustible
loading
and duration of design
basis fires.
During discussions
the licensee
agreed
to compare
the criteria for fire hazard
evaluation listed
by Branch Technical Position with the criteria
used
by the licensee
and submit this comparison
to the
HRC with
the evaluations
of changes
to the program discussed
above.
(3)
Some evaluations
did not provide sufficient information to
allow review by a person
who is not familiar with the plant.
To give
a specific example,
Fire Protection
Program
Change
Evaluation
FHARE Ho.
25 discusses
the issue of separation
between
redundant trains
for the centrifugal charging
pumps.
The
NRC granted
an exemption
request
to not provide
a
3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire barrier between
the
pumps
because
it was stated
there
was
a
50 foot separation
between
undampered
duct
between
the fire areas.
The licensee later determined
that the distance
was actually about
30 linear feet.
The evaluation
does
not specifically address
the issue that the probability of
malfunction of equipment
important to safety
may be increased.
The
evaluation
concludes
that the configuration is acceptable
as is based
on available fire suppression
and detection,
and combustible
loading
(fire severity)
less
than
30 minutes
and
60 minutes for the areas.
The inspector considers
that if a fire initiates in this area,
there
is
a greater likelihood of loss for a 30 foot vice
a 50 foot separation
between trains.
The shorter distance
allows less dissipation of
heat
and
smoke,
which can increase
the probability of a malfunction
of safety equipment.
The significance of this increased
probability
cf >alfunction>>ill be assessed
by detailed
VRC review.
However, it
is
an increased
probability of a safety
system malfunction and should
be addressed
according to
In addition, the assumption
of 30 and
60 minute fires is based
on licensee
computer database
calculations
which appear
to have different assumptions
than
CYCLED
9.5. I for combustible
loading.
Detailed
NRC review will assess
the
acceptability
of the licensee
combustible
loading calculations with
respect
to evaluations of fire hazards.
Although an acceptable
level of fire protection
may be the
preliminary
NRC conclusion,
a detailed
NRC fire protection review is
required to confirm this conclusion.
In a meeting
on January
17,
1990 with NRC Region
V, the licensee
stated that each of the
descrepancies
would be evaluated
according to Nuclear Safety Analysis
Center guidelines for 10 CFR 50.59 evaluations.
These evaluations
will be submitted to the
NRC for review.
The licensee
stated
these
evaluations
would be submitted
by April 15,
1990.
After reviewing
the submittal,
the
KRC will issue
a Safety Fvaluation Report.
In
this manner,
a detailed
NRC evaluation will be completed,
and the
approved fire protection
program will become consistent with the
as-built plant
~
According to tie license condition,
changes
to approved
program
features
which decrease
the level of fire protection should
be
made
with an application for license
amendment
pursuant
to
At least
15 of these
discrepancies
appear
to reduce
the level of
fire protection provided
by features
documented
in the
VRC approved
program.
Because fire supression, fire watches
and other
compensating
measures
are provided,
the inspectors
preliminarily
conclude
t,hat safety in the plant has not been
reduced to an
unacceptable
level.
However, these
types of changes
appear to
reduce
the margin cf safety
and increase
the probabilistic risk to
the plart.
The compensating fire protection
measures
(fire watches)
which the
licensee
is implementing are allowed
by
NRC fire protection
'egulations,
and are considered
by the
NRC to ensure
acceptable fire
protection to compensate
for the descrepancies
between the approved
program
and
as built plant.
However, the
NRC considers
these
measures
acceptable
interim corrective
measures
only.
Since these
discrepancies
have
been identified by the licensee
as early as
1982,
these
compensating
measures
appear to have
been
implemented
over the
longer term.
Therefore, this issue
should
be resolved promptly
since the licensee
should not continue to rely on short term
measures
to fulfill long term requirements.
The length of time between initial identification of this issue
by
the
NRC (6/10/87)
and this inspection
(12/4/89) is significant.
The
licensee
has
been
aware the
NRC is concerned
that this issue
has not
beer resolved.
The inspector
reviewed
5 internal
PGSE memoranda
issued
over
a
15 month period in which
PGSE fire protection
personnel
requested
PGSE engineering
and licensing staff to
determine appropriate
resolution
and notification of the
hRC.
These
memoranda
stated that regulatory issues
may
be involved,
and that
n
e
o
Ircnpt resolution
was desired.
However,
no communication
was
made
wit,h the
NRC to resolve this issue.
The timeliness of corrective
action will be evaluated after technical
resolution of this item is
complete.
B.
(Closed)
Unresolved
Item 50-275/89-17-02,
Out of Service Positive
cb ji!j5 j lllla i
ll d
d << t~e1
Charoiiig
Punrps.
The
FSAR requires
the positive displacement
charging
pump
(PDP) to
be available for safe
shutdown in the event afire renders
the
centrifugal charging
pumps inoperable.
Inspection report 89-17
identified that the
POPs in both units
had
been out of service for
extended
periods during the previous
two years.
Furthermore,
inspection report 89-17 noted that the
pumps
had each
been
continuously out of service for over three
weeks during calendar
year
1989.
( 1)
As corrective action for lack of PDP operability controls,
the
inspector verified that the licensee
implemented administrative
controls to ensure that the
PDP was returned to operability ir
one week,
and that the centrifugal charging
pump rooms
had
operable fire detection
and supression
systems, fire watches,
limited fire loading
and limited hot work.
Justification for
Cortinued Operation
89-13
was
issued
which increased
restrictions
on operation
and maintenance
as
a result of PDP
inoperability,
The lack of administrative controls to ensure operability of
the
PDP, which is required for safe shutdown, is considered
a
violation of Technical Specification 6.8. 1, which requires
implerpentation
and use of procedures
for
RG 1.33, Appendix
A
items, which includes
This is
considered
a violation (50-275/89-33-01).
As
a corrective
action,
the licensee
changed
the charging
system operatino
procedure
to require
a
7 day limit on the time the
PDP can
be
out of service during plant operation, after which, plant
shutdown must
be initiated.
The inspector verified licensee
corrective action for this violation, and considered it
adequate.
(2)
The inspector
reviewed the licensee's
10 CFR 50.59 evaluation
for the inoperable
PDP,,and
found it did not completely address
the question of whether or not there is
a possibility of a
malfunction of a type different than that evaluated
in the
safety analysis report.
The licensee
stated
"There is no
change
in the configuration of either Unit that would create
.the possibility for an accident or malfunction of a different
type thar evaluated
previously in the safety analysis report."
However, the inspector
found that the
FSAR fi.re hazards
analysis
(Section 9) requires
the
PDP as
a backup in the event
of a fire ir, the charging
pump room.
This indicates that the
associated
analysis
assumed
the
PDP would be available to
ma..'ntain
RCS inventory.
Secause
the
FSAR fire hazard analysis
C
r
requires
use of the
POP
as
a backup,
shutdown of the plant
during
a fire without use of a charging
pump was not adoressed
in the.
FSAR fire hazard analysis.
In response
to this
particular question,
the licensee
should
have noted that
operation without
PDP to maintain
RCS inventory may be outside
the requirements
of the Fire Hazard Safety Analysis.
The
licensee
did not address
maintaining
RCS inventory in the 50.59
evaluation.
however,
in the associated justification for
continued operation,
the licensee
stated that shutdown, if
required,
would be implemented
according to the plant procedure
"Loss of All Charging".
The inspector considers
that the
procedure
"Loss of All Charging" meets
the safety analysis
requirement
concerning
RCS inventory since,
according to the
licensee,
the procedure
should allow a safe
shutdown while
maintaining
RCS inventory such that pressurizer
level remains
in the indicating band
as required
Section L.2.:h.
The licensee
(P.
Kao) agreed
to attach this
50.59 review to the April 15 submittal for further
NRC
evaluation to determine if the evaluation properly addressed
%he criteria of 10 CFR 50.59.
Gereric Letter
(GL) 81-12, which specifically addressed
operability requirements
for safe
shutdown equipment,
was
addressed
t.o plar.ts licensed
before January
1, 1979.
Diablo
Canyon >>as
not. licensed
at that time, but had
been
issued
a
construction permit,
had received
and
was
participating in resolutior: of NPC fire protection requirements
in other areas.
A cursory review of this letter would have
highlighted these
weakness
in the Diablo Canyon program.
However, except for equipment already
addressed
in Technical
Specificat:ions for other purposes,
the licensee
has not
provided. these administrat.ive controls.
The licensee
apparently did not recognize
the current lack of administrative
controls covering safe
shutdown
equipment operability.
To
date,
there are administrative controls to ensure
equipment
operability only for equipment already controlled by Technical
Specifications (for reasons
other than safe
shutdown during
a
fire)', and for the
PDP
as
a result of this specific issue.
As one of the 'corrective actions associated
with the lack of
administrative controls for safe
shutdown equipment,
the
licensee
is evaluating
the type of additional administrative
controls required for safe
shutdown equipment.
The inspector
reviewed the applicable
NCR reports
and minutes of Technical
Review Group meetings,
and attended
a Technical
Review Group
meeting concerning this issue.
The licensee
has characterized
safe
shutdown equipment
operability controls
as follows:
equipment controlled
by
Technical Specifications,
safety related
equipment,
and
non
safety related
equipment.
(a.) Technica) SPecificatiun
Equuiment
The licensee
concluded that equipmient controlled
by
Technical Specifications
requires
no additional controls.
For example,
Technical Specifications
require operability
of the source
range
nuclear instrumentation
in all modes
but
Y(ode I.
Based
on equipment history, the licensee
found that, although significant periods of maintenance
have 'occurred, at least
one source
range instrument
has
been operable
during Yode 1.
This operability is prudent,
since the change
from Vaode
1 to Viode 3 can
be unscheduled.
The licensee
assessment
appears
to be adequate.
(b. ) ~Safet
Related
EquiPaient
The licensee
found that operability of some safety related
equipment is not directly controlled,
however,
administrative maintenance
procedures
place
a high
priority on returning safety related
equipment to,service.
Eased
on
a review of equipment history which showed that
all inoperable safety related
equipment
was promptly
returned to service,
the licensee
concluded that the
existing ad()instrative controls requiring
a high priority
fc> n:airitenance of safety related
equipment already
ensure
acceptable
operability of safety related
equipment
requirec for safe
shutdown.
This assessment
appears
to be
adequate,
(c.)
Nnr Sujet
Related
Equipment
The licensee
determined that several
items required for
safe
shutdown were
non safety related.
The operability
records nf the specific valves in the
ASW and
CC'W systems
were reviewed,
and the licensee
determined that there
had
been ro significant periods of inoperability.
In
addition,
t.he safe
shutdown function of several
valves is
to remain shut
and not change position.
Therefore, if the
valve is made inoperable for maintenance
on the operator,
its safe
shutdown function is not impaired.
The licensee
concluded,
however, that based
on the lower maintenance
priority of non safety related
equipment
and experience
with the
PDP, operability controls
may be required for non
safety related
equipment.
Based
on the review of the operability records,
the licensee
.determined that, with the exception of the
PDP, there are
no
indications to date that required safe
shutdown equipment
would
not have
been operable.
At the time of this report the
licensee
is considering
what type of administrative controls
would be most appropriate
to implement for non safety related
equipment to ensure operability of items required for safe
shutdown.
Options under consideration
are procedural
controls,
inclusion of requirements
in the maintenance
of database
to
automatically include operability requirements
in wort,
e
planning,
and other met.hods.
Based
on the licensee's
correc tiv<< action to da'te
and tracking of this item in the
licerse<<'s
NCP ard Techriical
Review Group process,
this iteo. is
closed.
(Closed)
Unresolved
Item 50-275/88-02-05,
Calibration of Resu1atory
Fute
1.97 Instrumentation.
An earTier
NRC inspection
observed
tVat
calibration intervals for some Regulatory
Guide
(RG) 1.97
instrumentation
were not documented.
Specifically Unit 2
4
kV bus
I'oltage,
Unit 2 bus
F voltage, Unit 2 battery 21, 22,
and
23 voltage
and current
hao
no calibration recorded.
Also, calibrations for
several
other
RG 1.97 instruments
showed that they were past the
three year calibration interval
and therefore
overdue for
calibration; specifically Unit
1 battery
11, 12,
and
13 voltage
and
currcrst.
After a review nf all
PG 1.97 instruments,
the licensee
determined
that calibration intervals for these
and
12 other instruments
hac
'not been entered
into the Recurring
Task Schedule
(RTS) after
irstallation.
The licensee verified that calibration of battery
11,
12,
13s Pl,
22 and
23 voltage
and current indicators
were overdue.
The battery voltage
and current instrumentation calibrations
were
due 6/30/88 ard performed
between 7/7/88 ard 7/12/88.
Five of these,
irstruoerts
were found in tolerance,
the sixth was out of tolerance
hy 5i of scale,
which is more than the
2X tolerance
allowed.
)
Tt e licensee
entereo
the missing instrumentation calibration tasks
into -the
RTS and performed calibrations
as required.
This item is
closed.
Since licensee
procedure
AP C-450 "Preventative
h:aintenance
Program"
requires
RG 1.97 instrument calibration intervals to be entered
in
the
RTS, the fact that these
required preventative
maintenance
tasks
were
not. included in the
RTS is
a violation of AP C-450 and
therefore
a violation of NRC requirements.
The violation is
considered
a
non cited violation in accordance
with 10 CFR 2,
Appendix C, V.G. 1 (50-275/89-33-02).
XOpen) Unresolved
Item 50-275/88-02-01,
Resistor
Networks
Used
as
Isolators
'the .licensee
uses resrstor
netwo~rs
instead of~iso ation
~amp ifsers
to isolate
some
Class
IE instrument signals
from Class
2
systems.
The licensee
stated that computer analyses
supported
satisfactory resistor
network performance
as isolators,
but that
no
test
had
been
performed
on these isolation devices.
Design
and analysis
information for RCS pressure
and steam generator
wide range level indication resistor
networks
has
been provided to
the
NRC staff for further analysis.
This item will remain
open
pending conclusion of the
NRC evaluation.
tOP
),8
7
7
- 8-7,75/88.))2- 2,
Rtt
.87
R 8
7
~d
and Diversity of Stean(Generator
Level Indication.
A common power
inverter (Division IV) supplies
power to all four wide range
steam
generator
level indication channels.
Deviation SS-02-02
was issued
because
a single
power source
does
not meet
RG 1.97 diversity
requi'remehts.
The licensee
stated that this configuration met the
requirements
cf PG 1.97
because
rarrow range level
inoication
and auxiliary feedwater flow are
powered
by diverse
electrical
busses.
NRC inspection report 88-02 noted that steam
generator
level appears
to fall below the range of narrow range
indication during several
accident scenarios.
This issue is under technical
review by the
NRC staff and will
remain open.
(0 en) Unresolved
Item 50-275/88-02-03,
Lack of Recorders
Used
for'st
Accicent Iionitorin9'for Neutron 'Flux Indication.
neutron f1ux
is identified as
a Type AC,ategory
1 varia~ew ic
must
be
monitored in the post accident
phase.
RG 1.97 requires that this
variable
be recorded.
The licensee
submittals
to the
NRC for
RG 1.97 variables listed
neutron flux as continuously indicated,
but not recorded.
However,
this discrepancy with
was not identified as
such.
This
sub> ittal is under
KRC staff review.
Therefore, this item remains
open.
(Closed)
Fol1ow-up Item 50-275/89-17-01,
Inade9uate
Procedural
Ruida<ice to Ensure
Fire Protection.
Desi~nReview.
As
a resuit of
inacequaYe
design review, 6o fire protection water supply system
valves
were not verified operable
upon completion of a modification.
This was
documented
in LER No. 323/89-03.
The licensee
determined
the cause
to be inadequate
procedural
guidance
and insufficient
communication
between
engineers.
The licensee
took corrective
action tc revise or issue
procedures,
and to train individuals.
To evaluate
licersee corrective action,
the inspector
reviewed
procedure
AP C-1SS,
"Design
Change Operability Testing Program",
which was
issued
to specify the review of design
changes
with
respect
to operability testing requirements.
This procedure
required specific interaction
between
various engineering
and
operational
groups to ensure
an adequate
scope of review for design
change operability tests.
This procedure
appeared
adequate.
'The inspector
reviewed procedure
AP C-1S1,
"Onsite Plant
Yodification Administrative" which was revised to expand
the design
change
sponsor's
scope of review and required
increased
interaction
between
various engineering disciplines
and plant departments.
The
inspector also observed that design
change
sponsors
are specifically
required
by procedure
3e60N "Operating Nuclear
Power Plant Design
Change" instructions
and checklist to review fire protection
and
other interface requirements
for design
changes.
Basec
c n this review and discussions
with plant personnel
involved
in the desigr:
change
process,
this corrective action appeared
adequate.
This item is closed.
S
4
(Closed)
Follnwug
1 tap: 50-275/59-17-03,
Del~aed Notification of
Ciffsite Fire Dejartment.
During
a wilcl'land fire, noti7fcatson of
~tie California Bepartnient of Forestry
was delayed
10 to
15 minutes
because
the sheriff's of'fice followed "Unusual
Event" procedures
to
notify other state
agencies
before notifying the California
Department of Forestry to respond to the fire.
Also, during the
event,
the shift supervisor
incorrectly assumed
that the
NRC
resident. inspector would provide updates
of the event to the
NRC
Headquarters
Operation Center.
The inspector
reviewed the licensee
procedure
Y<-6 "Emergency
Procedure,
Non Radiological Fire," which was revised to add
a step
requiring the shift foreman to "immediately notify the California
Department of Forestry,
San Luis Obispo County Fire 'by calling 911
or by using
CDF radio/telephone."
This resolves
the previous lack
of prompt, direct notification of offsite fire fighting assistance.
Concerning
NRC notifications during events,
the inspector
observed
, that this and other
emergency
procedures
require licensee
personnel
tc notify NRC Headquarters
directly.
The assumption
that the
NRC
resioent
inspector woulo provide updates
to
NRC Headquarters
appears
to have
been
an isolated
personnel
error.
Licensee corrective
action included training of operators
to remind them that
NRC
residents
do not provide updates
to
NRC Headquarters.
Based
on
these corrective actions, this item is closed.
(Closed)
Follow-up Item 50-275/89-17-04,
Corrosion of the Fire h'ater
System~Pi
in
.
7he licensee identified col rosion pr'oducts" in the
turbine building fire water system piping.
Immediate tests
to
identify pipe wall thinning and sprinkler head plugging showed that
the corrosion
and resulting particulate
had not impaired fire
protection.
These tests
and evaluations
appeared
adequate.
The
licensee initiated an enoineering
evaluation to determine
the long
term effects of'his corrosion
and possible
prevention or mitigation
of',he corrosion.
The inspector
reviewed the licensee
analysis
and assessment
of
corrective actions,
and considers it adequate.
The analysis
.
determined
dissolved
oxygen to be the cause of corrosion.
The
licensee
is evaluating options of corrosion inhibitors, changes
in
pipe flushing schedules,
and other means of reducing dissolved
oxygen concentrations.
Based
on licensee
actions to date
and
tracking of this item in the licensee
NCR and commitment tracking
system, this item is closed.
(Closed) Follows Item 50-275/89-17-05
Sealiny of Fire Barrier
Generic Letters 88-04, 88-56
and 89-52 require the
licensee
to evaluate
the seals
on.dampers,
doors
and sealing
materials
to identify sealing
problems.
The inspecto~
found that the licensee
is performing these
evaluations
and -investigating various
means to ensure
adequate
seals.
These evaluations
range
from damper test
and observation
programs to the option to temporarily shut
down ventilation if
0
10
dan f ers,
can
not
be closed against air flow.
Based
on adequate
'icersee
ef ort tn date
and the fact that this iten; is being tracked
bp licensee
commi tnients, this i teni is closed.
(Closed)
Fol I os;-up
Item 50-323/89-02-01,
Three Annunciator l<indows
are lfecessary"for
llnambi9uous Accident
$fitigat on S~ste~ Actuatino
Circuitry
Af~SAC) Indication.
The control
room annunciator
indication for armed
ANSAC is an input into the
AViSAC trouble windov:
signals.
Therefore,
the control
room operators
can not quickly and
accurately
diagnose
an
ANSAC trip without relying on other
information.
To correct the ambiguous indication, the licensee will add
a third
AYiSAC annunciator
window labeled
"ANSAC trip," which provides
unambiguous
indication of an
ANSAC trip.
The inspector
observed
that this third window has
been installed in Unit 1, and is
scheduled
to be installed ir, Unit 2 during outage
2P3 (Yarch,
1990).
The third wirdov; is schedul'ed
to be installed in the simulator
by
Februarv,
'990.
These
schedules
are
documented
in the licensee
act,ior. requests
and conmitment tracking systems.
Based
on
irsiallation of the
"AYiSAC trip" window in Unit
1 and documentation
<<f f'lans for t.hese installations
in Unit 2 and in the simulator,
tt:is iten is closed.
(Closed)
Follow-uy Item 50-323/89-02-02,
Separation
of AYiSAC from
.ttte!eactor
Protection
System~(KPS
.
Yhe 1icensee
has
not cousuitted
tc me&. t.lie electrical
separation
requirements
of R6 1.75.
However,
the separation
between
the
AVSAC input signal wiring (steam
generator
low level
and turbine impulse pressure)
and the Class
1E
wiring ir. the
RPS analog
process
cabinet
was determined
to be less
tl>an desirable.
In each
RPS channel
analog cabinet,
Class
2 spare
AVSAC wires
as well as the Class
2
AhlSAC signal wires from the
isoliation aniplifier v,ere bundled with Class
The
AYiSAC spare viires bundled with signal wires from the four RPS
cabiriets entered
the
AhlSAC logic cabinet,
where
FSAR Class
1E
separat,ion criteria were followed for the signal wire.
However, all
four channeIs
of the spare wires, which were taped at both ends,
were coiled in the lower section of the
ANSAC cabinet.
In this
configuration,
a failure in the
ANSAC non-class
1E cabinet could
negate protective actions
because
of lack of physical
separation
between
the output of isolators
and Class
IE wiring.
However, this
possibility is small considering that the highest potential in the
AViSAC cabinet is
120V AC for processor
power supplies,
and the rest
of the
ANSAC circuitry is of much lower energy.
The inspector
noted that the Unit
1
ANSAC cabinets
had
been modified
to have
no spare signal wires.
The
ANSAC Class
2 signal wires in
the
RPS cabinets
are bundled with RPS Class
1E signal wiring,
however,
the
AVSAC wiring was wrapped with varglass,
which meets
the
FSAR physical
separation criterion.
The input wires from all
4
charnels
were bundled together in the
AYiSAC logic cabinet,
however,
the individual input wires were wrapped with varglass.
Although 5
inch separatior,
betv;een
each
channel
would be more desirable,
the
insulation meets
the separation criteria of the
FSAR.
In addition,
the
lov, energy of the circuit.s in the
AViSAC cabinets
reduces
the
~\\,
'ikelihood of a problem.
The Unit 2
AYiSAC cabinets
are scheduled
to
be tncdifie<< in the
same
manner during outage
2R3 (March 1990).
The
licensee
is tracking the comlmitment to complete this design
chanoe.
This item is closed.
M.
(Closed)
Follow-up Item 50-323/89-02-03,
AMSAC Testing
Procedures
ffo~ fh'rSiEL't power testing
and refue'ling ooutage end-to-,end,testino
were not finalized.
Also, the test requirements
had not been
incorporated
in the licensee
Recurring Task Schedule.
The inspector
reviewed procedure
STP-I-92A, Surveillance Test
Procedure
AYiSAC functional test.
It appeared
adequate.
The
inspector
found that at power surveillance test procedures
for Unit
1 had
been
approved
and scheduled.
Unit 2 procedures
had
been
prepared,
but could not be scheduled until the end of the outage
(April, 1990).
Preparation of refueling
outage
end-to-end testing
and entry
o
tesi
requirements
int<< the Recurring
Task Schedule
were
ir. progress for Unit
1 and scheduled for Unit 2.
The inspector
verified that
PGSE letter DCL-88-049 committed the licensee
to
quarterly at power testing of AYSAC, end to end (refueling outage)
testini<< every
18 months,
and entry of the surveillances
in the
Recurrir 9 Ta
k Schedule.
Based
an these
observations
and licensee
cottmitmerts, this iten is closed.
7.
Exit hie~etin
An exit meeting
was held with the licensee staff cn December 8,
1989.
The
specific concerns
addressed
in this report were discussed
with the
licensee
during the above meetings
and were acknowledged
by the licersee.
0
L,
9
I