ML16341G635
| ML16341G635 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 07/07/1992 |
| From: | Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML16341G634 | List: |
| References | |
| 50-275-92-16, 50-323-92-16, NUDOCS 9207270296 | |
| Download: ML16341G635 (48) | |
See also: IR 05000275/1992016
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION V
Report Nos:
Docket Nos:
License
Nos:
Licensee:
Facility Name:
Inspection at:
50-275/92-16
and 50-323/92-16
50-275
and 50-323
DPR-80 and DN-82
Pacific
Gas
and Electric Company
77 Beale Street,
Room 1451
San Francisco,
California 94106
Diablo Canyon Units
1 and
2
Diablo Canyon Site,
San Luis Obispo County,
Inspection
Conducted:
April 21 through June
1,
1992
Inspectors:
P. Morrill, Senior Resident
Inspector
M. Miller, Resident
Inspector
Approved by:
Summary:
P.
H.
ohnson,
Chief
Reac
r Projects
Section
1
~(~(~~
Date Signed
Ins ection
on
A ril 21 throu
h June
1
1992
Re ort Nos.
50-2
5 92-16
and
50-323 92-16
1
t d:
t tt
t
p tt
1 pt
t
p
tt
,
1 t
d
surveillance activities, followup of onsite events,
open items,
and licensee
event reports
(LERs),
as well as selected
independent
inspection activities.
Inspection
Procedures
37051,
50073,
61726,
62703,
71707,
71710,
92700,
92703,
93702,
and 94600 were used
as guidance during this inspection.
Safet
Issues
Mana ement
S stem
SINS
Items:
None
Results
General
Conclusions
on Stren ths
and Weaknesses
Strengths:
A more conservative
approach to the scheduling of Instrumentation
and
Control surveillances
was noted.
This involved
a licensee initiative to
discontinue routine use of the
25X grace period allowed by the Technical
Specifications for the performance of surveillances
(paragraph 6.a).
9207270296
920707
ADOCK 05000275
8
i
Meaknesses:
The
NRC identified several
problems which had not been identified or
appropriately
documented
by the licensee.
These included the installa-
tion of plastic tarps in front of Appendix
R lighting (paragraph
S.a)
and
the violations identified below.
Si nificant Safet
Matters:
None
Summar
of Violations:
One Severity Level
IV violation was identified,
involving the loading of two one-ton chainfalls to approximately
2400
pounds
each wile installing
a resin cask lid (paragraph 3.b(2)).
Two
non-cited violations were also identified, involving inconsistent
fastener torques for two residual
heat
removal
(RHR) system valves
(paragraph
S.b)
and failure to declare
a Unit 2 main steam safety valve
inoperable after test equipment
used to establish its setpoint
was found
to have failed (paragraph
10.k).
en Items
Summar
Three
new followup items were opened.
Ten followup items were reviewed;
9 were closed, I remains
open.
Twenty LERs were reviewed;
19 were
closed,
1 remains
open.
Persons
Contacted
acific Gas
and Electric
Com an
G. H. Rueger,
Senior Vice President
and
General'anager,'uclear
Power Generation
Business
Unit
J.
D. Townsend,
Vice President
and Plant Manager,
Diablo Canyon
Operations
W. H. Fujimoto, Vice President,
Nuclear Technical
Services
D. B. Miklush, Manager,
Operations
Services
- H. J. Angus,
Manager,
Technical
Services
- B. W. Giffin, Manager,
Maintenance
Services
- W. G. Crockett,
Manager,
Support Services
~J.
E. Molden, Instrumentation
and Controls Director
- W. D. Barkhuff, guality Control Director
- R. P.
Powers,
Mechanical
Maintenance Director
T. L. Grebel,
Regulatory Compliance Supervisor
H. J. Phillips, Electrical Maintenance Director
- J. A. Shoulders,
Onsite Project Engineer
- S. R. Fridley, Operations Director
- R. Gray, Radiation Protection Director
J. J. Griffin, Senior Engineer,
Regulatory Compliance
J.
V. Boots,
Chemistry Director
- T. A. Moulia, Assistant to Vice President,
Diablo Canyon Operations
- R. Kohout, Safety,
Health
and
Emergency Services Director
- D. P. Sisk, Regulatory Compliance
Engineer
D.
R. Stermer,
Power Production
Engineer
M. R. Tresler,
Project Engineer
R. Clark, Assistant Project Engineer
R. Gagne, Acting Radwaste
Foreman
U. A. Farradj, Fire Protection
Engineer
R. A. Waltos, Senior
Power Production
Engineer
S.
F. Shrefler,
Mechanical
Maintenance
Engineer
B. D. Pogue,
System Engineer
R. Ortega,
System Engineer
R. Watson, guality Assurance
Engineer
San Luis Obis
o Count
Officials
R. Hendrix, Administrator
L. Williams, Deputy Administrator
V. Horici, Emergency
Response
Coordinator
- Denotes those attending the exit interview.
The inspectors
interviewed several
other licensee
employees
including
shift supervisors,
shift foremen
(SFH), reactor
and auxiliary operat'ors,
maintenance
personnel,
plant technicians
and engineers,
and quality
assurance
personnel.
2.
0 erational
Status of Diablo Can
on Units I and
2
Dunng the )nspectson
period,
Unst I operated
at
IOOX power, except for
April 25-27,
and Hay 2-3,
1992.
On April 25, Unit I reduced
power to 50X
for routine condenser
cleaning.
While conducting maintenance
on the l-l
main feedwater
pump,
vacuum in the condenser
was lost, causing the main
turbine
and the reactor to trip.
Unit I was restarted
on April 26 and
was at full power on April 27.
This .event is described in paragraph
4.a
below.
During the period Hay 2-3,
1992, Unit I was curtailed to 50X
power for routine condenser
cleaning.
Unit 2 operated
at
100X power during the reporting period.
3.
0 erational
Safet
Verification
71
07
a.
General
During the inspection period, the inspectors
observed
and examined
activities to verify the operational
safety of the licensee's
facility.
The observations
and examinations of those activities
were conducted
on a daily, weekly or monthly basis.
On a daily basis,
the inspectors
observed control
room activities to
verify compliance with selected
Limiting Conditions for Operation
(LCOs)
as prescribed
in the facility Technical Specifications
(TS).
Logs, instrumentation,
recorder traces,
and other operational
records
were examined to obtain information on plant conditions
and
to evaluate trends.
This operational
information was then evaluated
to determine whether regulatory requirements
were satisfied.
Shift
turnovers
were observed
on
a sample basis to verify that all perti-
nent information on plant status
was relayed to the oncoming crew.
During each
week, the inspectors
toured accessible
areas of the
facility to observe the following:
(I)
General
plant and equipment conditions
(2)
Fire hazards
and fire fighting equipment
(3)
Conduct of selected activities for compliance with the
licensee's
administrative controls
and approved
procedures
(4)
Interiors of electrical
and control panels
(5)
Plant housekeeping
and cleanliness
(6)
Engineered
safety features
equipment alignment
and conditions
(7)
Storage of pressurized
gas bottles
The inspectors
talked with control
room operator s and other plant
personnel.
The discussions
centered
on pertinent topics of general
plant conditions,
procedures,
security, training,
and other aspects
of the work activities.
-3-
The inspectors
also
accompanied
auxiliary operators
during their
rounds in the plant and reviewed the associated
log entries.
On two
occasions
the inspector verified that auxiliary operators
were
completing their rounds consistent with the -log sheet entries.
No
instances
of improper log entries
were identified.
b.
Radiolo ical Protection
(1)
The inspectors periodically observed radiological protection
practices to determine
whether the licensee's
program was being
implemented in conformance with facility policies 'and proce-
dures
and in compliance with regulatory requirements.
The
inspectors verified that health physics supervisors
and
professionals
conducted
frequent plant tours to observe
activities in progress
and were aware of signi'ficant plant
activities, particularly those related to radiological
conditions and/or challenges.
ALARA considerations
were found
to be
an integral part of each
RWP (Radiation
Work Permit).
(2)
Radioactive
Waste Shi ment on Ha
28-29
1992
The inspectors
observed
the licensee's
preparation to ship
a
high integrity container
(HIC) of radioactive resin to
Richland,
Washington for disposal.
The licensee
planned to
survey
and
remove the HIC from storage
in a permanent
plant
shield cask
and place the HIC in a transportation
cask carried
on
a sole-use
flatbed truck.
A shielded
boom crane
was
used
for this movement.
The evolution was significant in that it
occurs infrequently
(1 to 2 times
a year at most)
and the
had
a surface
dose rate of over 30 R/Hr.
The inspectors
observed
the licensee's
pre-brief meeting which covered
communications,
surveys,
exclusion areas,
ALARA considerations,
crane operations,
and contingency plans.
The inspectors
observed
the HIC transfer,
which proceeded
without incident.
However,
when placing the lid on the
shipping cask,
licensee
personnel
were initially unable to get
it to seat properly.
The lid was supported
by three slings
and
restrained horizontally with respect to the cask by two guide
pins.
With the lid resting
on the top of the cask,
the
licensee's
riggers replaced
two of the slings with chainfalls.
This allowed the lid to be leveled
by an individual standing
on
top of the lid and adjusting the chainfalls.
The rigger
leveling the lid was shielded
from the
HIC by the lid itself.
The lid was approximately
3 inches higher than the seated
position during the evolution.
After several tries the lid
seated
and the rigging equipment
was removed.
During this evolution, the inspector questioned
licensee
personnel
as to whether the chainfalls being used
had adequate
capacity.
The licensee
personnel
present
stated that they were
confident that the riggers
knew what they were doing.
The
inspector stated that the chainfalls looked like one-ton units
and the cask lid looked substantially heavier.
i
The next day,
Hay 29,
1992, the inspector conducted
a survey of
the shipping cask
on the truck.
Radiation levels
appeared
to
be, within regulatory limits and the cask appeared
to be proper-
ly secured
and labeled for shipment.
The inspector
observed
that the weight of the cask lid was approximately
7300 pounds.
The inspector discussed
the chainfall.capacity. with the Mecha-
nical Haintenance
Director
who determined that the chainfalls
used
had
been one-ton units.
The licensee's
quality evaluation
found them to have
a safe working load of 2000 pounds
each.
The load on each chainfall would have
been approximately
2400
pounds.
The inspector
noted that this was
a personnel
safety
issue
and that the individual standing
on the cask lid could
have
been seriously hurt if the chain
had parted.
The Manager
of Maintenance
Services
and Mechanical
Maintenance Director
agreed that
a mistake
had
been
made.
Licensee
personnel
stated
that when the next shipment
was
made the correct size chain-
falls would be available
and 'that their use would be required
by written instructions.
This,is an apparent violation of
and procedures for control of
rigging and load handling equipment.
(50-275/92-16-01)
The inspector
observed that this event involved riggers'and
contained
elements similar to the loss of offsite power event
in March 1991 which was caused
by a boom crane
under the Unit
1
500KV lines.
The Manager of Maintenance
Services
stated that
the
boom crane event in 1991
had
been
caused
by personnel
who
were not riggers,
but who had been trained to use the equip-
ment.
The inspector
acknowledged that two different groups of
personnel
were involved, but noted that both events
appeared
to
involve weakness
in the preplanning
and control of lifting or
rigging activities.
c.
Ph sical Secur't
Security activities were observed for conformance with regulatory
requirements,
the site security plan,
and administrative procedures,
including vehicle
and personnel
access
screening,
personnel
badging,
site security force manning,
compensatory
measures,
and protected
and vital area integrity.
Exterior lighting was checked during
backshift inspections.
No violations or deviations
were identified.
4.
Onsite Event Followu
93702
a ~
Tri
of Unit
1
Due to Loss of Vacuum on
A ril 25
1992
Between 9:00 p.m and 11:00 p.m.
on April 24,
1992, plant operators
decreased
Unit
1 power to approximately
50X to allow routine
cleaning of the east
water boxes.
Main feedwater
pump
(HFP) l-l was shut
down and cleared to make repairs to
steam'top
valves to the
MFP turbine.
At approximately 3:45 a.m;,
vacuum in the operating
west condenser
had degraded
to an absolute
pressure of approximately 4.5 inches
Hg.
The air in-leakage
which degraded
condenser
vacuum was apparently
due to leakage
through the large l-l MFP exhaust butterfly valve.
The on-shift crew attempted to stop the air in-leakage
by reseating
the closed
1-1
HFP exhaust
valve and by placing
a second air ejector
in service.
At 4:00 a.m., plant power was. being decreased
at
10
NW
per minute
and gland seal
steam
was re-established
to the
HFP 1-1
turbine.
The decrease
in power appeared
to be helping condenser
vacuum.
At 4:07 a.m., plan+operators
attempted to place the Nash
mechanical
vacuum
pump in service.
When the vacuum
pump suction
valve from the west main condenser
was opened,
backpressure
increased
rapidly and at 4:08 a.m.
a turbine trip and reactor trip
occurred.
The operators
subsequently
stabilized the plant in
accordance
with Emergency Operating
Procedures.
All safety related
equipment functioned
as expected.
At 5:10 a.m.,
the licensee
made
a
four hour non-emergency
report to the
NRC.
The inspector evaluated
the licensee's
investigation,
discussed
the
event with the on-shift crew,
and observed
the subsequent
plant
startup.
At approximately
1:00 p.m.
on April 27, after evaluation
of the event
and correction of the cause of the trip, the licensee
restarted
Unit 1.
The startup
was uneventful
and Unit
1 returned to
full power the morning of April 28,
1992.
The inspector
observed that the licensee's
investigation into the
trip identified the following information:
After the operators
had
begun to reduce
power at
10
HW per
minute,
condenser
vacuum was stabilizing.
The problem may have
corrected itself without further action.
Placing the
Na'sh mechanical
vacuum
pump in service
was
consistent with the licensee's
procedure
AP-7, "Loss of
Condenser
Vacuum".
However,
AP-7 did not contain the specific
steps
necessary
to place the Nash
pump in service.
The operators
in the turbine building at the Nash
pump did not
have
a local copy of the procedure to place the
pump into
service.
Due to the sense of urgency the operators
in the
control
room did not review the procedure to place the Nash
pump in service.
Unknown to the operators,
a seal
water isolation valve for the
Nash
pump was required to be opened
before the
pump could
operate properly.
When the suction valve to the west condenser
was opened, air entered
the condenser
through the
Nash
pump
seals.
The condenser
vacuum
pump suction check valve appears
to have
leaked severely in the reverse direction, allowing air flow
back to the condenser.
The inspector discussed
the event with the Operations Director to
determine the adequacy of the licensee's
corrective actions.
The
licensee
issued
an Incident Summary to alert other crews
and
added
precautions
to be taken in the procedure for shutdown
and clearing
of a main feedwater
pump.
The licensee
plans to inspect
each unit's
vacuum
pump suction line check valve during outages
1R5 and
2R5.
The licensee
also plans to complete
a review of all emergency
and
abnormal
operating
procedures
to identify situations in which opera-
.
tors might be dispatched
to perform equipment operation without
procedures.
By September
1,
1992, controlled local procedures
or
postings
are planned to be aided
based
on the review.
The inspector questioned
the adequacy of the licensee's
preparation
for taking
HFP l-l out of service.
Licensee
personnel
stated that
preparations
had been
adequate,
but that the plant had not responded
as expected.
This event
was described
in Licensee
Evqnt Report
(LER) 1-92-004,,issued
May 22,
1992.
The inspectors will followup
the licensee's
analysis of the event
and corrective actions during
review of the
LER.
No violations or deviations
were identified.
5.
Maintenance
62703
The inspectors
observed portions of,
and reviewed records
on, selected
maintenance activities to assure
compliance with approved
procedures,
Technical Specifications,
and appropriate
industry codes
and standards.
Furthermore,
the inspectors verified that maintenance activities were
performed
by qualified personnel,
in accordance
with fire protection
and
housekeeping
controls,
and that replacement
parts
were appropriately
certified.
These activities included:
Work Order No. C0096441,
BATP1, Replace
Mechanical
Seal
Work Order No. C0100168,10-142,
Cask,
Provide Support to Hove Cask
a.
Concerns
Raised
Durin
Observation of Boric Acid Transfer
Pum
aintenance
During an observation of Work Order
No C0096441,
BATPl, "Replace
Mechanical Seal," the inspector
noted the following concern:
(1)
When questioned
by the inspector,
the mechanics
performing
alignment of the
pump and motor coupling stated that guality
Control sign-off of the alignment would require removal of the
optical alignment tool used to align the coupling,
and instal-
lation of a double dial indicator,
an older, slower alignment
tool.
This was required
because
guality Control would not rely
on optical alignment information.
Further review by the inspector
and Mechanical
Haintenance
management
found that guality Control
had accepted
optical
alignment information a few months
ago,
but that the Mechanical
Maintenance
foreman
and mechanics
had not all been
informed.
Although both alignment methods
are valid, the inspector
was
concerned that the mechanics
and foreman were not informed of
processes
which could reduce radiation exposure
by reducing
time in radiation fields.
The licensee is reviewing training
and technical
information provided to foremen to determine if
timely and adequate
information is provided to Mechanical
Maintenance
foremen.
(2)
The work order required all workers to attend the tailboard
discussion.
However, tge mechanic performing alignment of the
pump and motor coupling had not attended
the tailb'oard.
The
licensee
stated that coupling alignment activities are highly
specialized,
and in this case did not require attendance
of the
tailboard, in that the mechanic
had
been briefed
on the job and
provided adequate
information to perform the work safely.
The
inspector
noted that the procedure
requirements
qnd expecta-
tions for specialists
regarding tailboard attendance
were not
specific,
and
may be subject to incorrect interpretation,
par-
ticularly during times of high work activity such
as
an outage.
The licensee
was planning to review expectations
regarding
attendance
at tailboards
before the next scheduled
outage.
These
issues will be followed during routine resident inspection
activities.
No violations or deviations
were identified.
~
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lj
6.
Surveillance
61726
By direct observation
and record review of selected
surveillance testing,
the inspectors
assessed
compliance with TS requirements
and plant
procedures.
The inspectors verified that test equipment
was calibrated,
and that test results
met acceptance
criteria or were appropriately
dispositioned.
These tests
included:
STP I-18W2D, Revision 5, Isotopic Calibration of Plant Vent Mid
Range
Iodine Radiation Monitor RH-32
STP R-3A, Use of Flux Mapping Equipment (Unit 2)
STP H-IIB, Measurement of Station Battery Voltage and Specific
Gravity (Unit I)
STP M-45B, Containment
Inspection
When Containment Integrity is
Established
(Unit 2)
STP H-51A, Routine Surveillance of Containment
Fan Cooler Unit
(CFCU) for Reverse
Rotation
(CFCU 2-4)
'a Q
I-18W2D
Revision
5
Isoto ic Calibration of
lant Vent Mid
Ran
e Iodine Radiation Monitor RM-32
During review of this surveillance activity, the licensee
noted that
several
Instrumentation
and Control
(I&C) calibrations
had
been
performed during the TS grace period
(25X beyond the surveillance
interval specified in the TS).
The inspector
was informed by IKC
technicians that routine use of the grace period was beiag stopped
by management,
and that surveillance activities have more recently
been
and will in the future continue to be performed within the
TS
inte>val.
This effort was documented
in problem identification
document
qE No. 9766, dated
Hay 27,
1992.
The inspector
reviewed
a
sample of recent
18C surveillances,
and found that more recent sur-
veillance tests
had been
accomplished within the
TS interval, rather
than during the grace period~
as
was the case for some surveillance
activities performed
more than six months
ago.
This licensee
initiative was noted
by the inspector to reflect
a more conservative
approach to the scheduling of surveillance activities.
b.
Pro
ram to
U
rade Radiatio
onitor Cal bration Procedures
The inspector
observed
implementation of a review of radiation
monitor calibration procedures
to identify vague or conflicting
guidance.
The licensee
had identified this effort in an action plan
and several
action requests.
The review and associated
effort
appeared
appropriate.
Followup of this effort will be performed
as
part of routine resident
inspection of surveillance activities.
No violations or deviations
were identified.
En ineerim
Safet
Feature Verificatio
71710
During the inspection period, selected
portions of the safety injection
system for Units
1 and
2 were inspected to verify that system configura-
tion, equipment condition, valve and electrical lineups,
and local
breaker positions were in accordance
with plant drawings
and Technical
Specifications.
No violations or deviations
were identified.
8.
Exam les of Inade uate
Problem Identif'cation
a.
lant Construction
Obscures
A
endix
R
Li htin
In Unit 1, Buses
G and
H switchgear
rooms, construction
work is
underway to strengthen
masonry walls.
In both rooms,
a plastic tarp
was
hung from floor to ceiling along the length of one wall to pro-
tect the switchgear
from debris
gener ated
by the constr uction work.
On Nay 19,
1992, the inspector
noted that the Appendix
R emergency
lighting for both rooms
was covered
by the tarps.
Emergency light-
ing is installed in the
4KV switchgear
rooms to provide lighting for
operator actions in the event of a fire.
Operator actions
are docu-
mented in procedures
OP-AP-8A and
B, Control
Room Inaccessibility,
and
EP H-10, Fire Protection of Safe
Shutdown Equipment.
The licensee
documented
the problem in AR No. 266808,
and replaced
an approximately
6 ft by 5 ft square
section of the plastic in front
nf the lights with transparent
plastic.
The licensee
was planning
to evaluate operability of the lighting, and assigned
system
engineering responsibility for Appendix
R lighting to the fire
protection
and emergency
services organization.
Appendix R,Section III.J, requires
emergency lighting in all areas
needed for operation of safe
shutdown equipment.
The operability of
the lighting, and root cause of the failure to identify and correct
the problem until noted
by the inspector, will be followed as
Unresolved
Item 50-275/92-16-02.
Contradictor
Tor ue Docume
ation for Residual
Heat
Removal
S stem Valves
During review of the evaluation section of Action Request
No.
230770,
which identified recent indication of leakage
from Unit
1
valve
RHR 8702, the inspector noted that
a second
problem was dis-
cussed
concerning evaluation of inconsistent
torque documentation.
Torques of 1500 and
2500 ft-lbs for bonnet fasteners
were listed in
separate
reference
documents.
This inconsistency
was not identified
in a separate
Action Request
as required,
and therefore
may not have
been
reviewed by guality Assurance.
After the inspector identified
the issue,
the licensee
acknowledged that the lack of a separate
problem identification document
was inappropriate,
and initiated
a
second Action Request
(267237) to identify the problem.
The inspector
was concerned that,
based
on inconsistent
information,
fasteners
with incorrect torque
may have
been installed in the
plant.
Licensee investigation determined that the valve manufac-
turer initially recommended
1500 ft-lbs, but had later offered
different valve trim designs.
The manufacturer
then specified
2500
ft-lbs torque because it bounded all design options for this valve
line,
and documented this value in later vendor manuals.
Recent
vendor correspondence
validated this assessment.
Installation
records
from 1973 for original design valves
showed torques of 1800
ft-lbs, which the licensee
considers
an acceptable
torque value.
Additionally, the licensee
stated that these
valves are specifically
inspected for boric acid leakage at least every outage during sur-
veillance test
STP R-BC, Containment
Malkdown for Evidence of Boric
Acid Leakage, to comply with the licensee's
response
to Generic Letter 88-05 concerning boric acid corrosion of fasteners.
Because
no leakage
has
been previously observed,
the licensee
concluded that
this is further assurance
that fastener
torque is adequate.
The licensee
stated that this issue
had not been identified in the
past
because
no work had
been performed
on the two val.ves of
concern,
RHR 8701
and 8702, since original installation.
The failure to promptly identify a problem adverse
to quality is
considered to be
a violation of 10 CFR 50, Appendix B, Section XVI.
Because
the torque inconsistencies
did not appear to be safety
significant,
and because
licensee corrective actions
were prompt,
this violation is not being cited because
the criteria specified in
Section V.A. of the Enforcement Policy were satisfied.
(NCV 92-16-03, 'Closed)
1
e
10
i
c.
Uncontrolled Information in Electrical
Panels
During routine inspection activities, the inspector
observed
uncontrolled,
outdated,
and in some cases
inaccurate
drawings
and
circuit lists attached to the inside of several electrical
breaker
panels.
The presence
of uncontrolled information in electrical
panels,
particularly in Cl.ass
1 panels,
was of concern.
After
discussion
with the inspector,
the licensee
determined that this
information was inconsistent with the requirements
of Administrative
Procedure
AP C-55,
and indicated lack of a formal program for
updating electrical labels
a% part of the design process.
Licensee
work control
and operations
procedures
requ'ire
use of controlled
drawings
and information for work associated
with these
panels.
Inspector observations
indicated that the existence of these
uncontrolled drawings did not appear to have
encour'aged
the licensee
to circumvent the use of controlled drawings.
Therefore,
the safety
significance of uncontrolled drawings in panels
appeared
low.
Removal of uncontrolled information from inside electrical
panels
will be followed by routine inspection activities.
One non-cited violation was identified (paragraph 8.b).
9.
Potential
Weakness
in Understandin
the Recircu at'o
P ase
Desi
n
s
s
In recent months,
the inspector
has noted the occurrence
of five issues
which appear to have
been partially caused
by insufficient awareness
of
design basis
requirements for the
ECCS recirculation phase.
They are
as
follows:
~
Diaphragm Valve Leakage
May Have Exceed Part
100 Limits (LER 50-323/
92-09).
~
Charging
Pump
and Safety Injection
Pump Runout Limits Exceeded
(NCR-
DC0-92-NS-N007).
~
Leakage
Path Vulnerability Via Volume Control Tank (VCT) Outlet
Check Valve (LER 50-275/92-01).
~
Inaccurate
Single Failure Analysis Allows the Possibility that
Com-
ponent Cooling Mater
(CCW) Heat
Load may Incr ease,
and Charging
Pump
Bearing Temperature
Limits may be Exceeded
(LER 50-275/91-18).
~
Inaccurate
Single Failure Analysis Allows the possibility that
Long
Term Containment
Temperature
Profile Assumptions
may be Exceeded
due
to Lack of Containment
Spray
(NCR DC0-92-EN-N002).
The inspector identified to the licensee that these
issues
may indicate
insufficient understanding
and integration of ECCS recirculation
phase
design requirements,
with the potential for additional safety issues
which have not yet been identified.
The licensee
agreed to consider
and
evaluate this possibility.
The licensee identified that this may be
a
generic Westinghouse
design issue,
and was planning to explore this issue
with the Westinghouse Owners'roup.
Followup of the licensee's
evaluation will be tracked
as Followup Item 50-275/92-16-04.
~
~
10.
Licensee
Event
Re ort Fol
owu
92700
a.
Unit I Control
Room Ventilation
S stem Outs'de
esi
as's
LER 50-275 83-39
Revision
0
C osed
On November 21,
1991, the licensee identified that the control
room
ventilation system could be outside its design basis
as
a result of
a single failure of a booster fan or its associated
The
fans are positioned in parallel flow paths,
and failure would allow
unfiltered air into the consol
room during the pressurization
mode.
As a temporary corrective measure,
the licensee
issued
Operations
Department night orders to the control
room discussing
the vulnera-
bility, and installed flow indicators
on ducts in the control
room.
The licensee
also changed
emergency
operating
procedures
to require
verification that flow is occurring in the proper direction through
the ducts
by observing the flow indicators.
The licensee
has prepared
and scheduled
a design
change to provide
permanent resolution of the vulnerability.
Therefore, this item is
closed.
b.
Unit I Check Valve Inservice Testin
eficiencies
R 50-2
5 84-4
evisio
Closed
!
C.
As a result of deficiencies
in the check valve inservice testing
(IST) program,
two additional valves were identified which require
testing in the closed direction.
Identification of these
valves
occurred after review of a Westinghouse
Information Letter dated
August 4,
1989.
The valves,
HS-5166
and -5177,
are in the steam
supply lines to the turbine driven auxiliary feedwater
(AFW) pump.
Testing
was in place to verify the valves
opened to supply steam
when required,
but not that the valves closed in order to prevent
an
intact steam generator
from blowing down through the
AFW pump steam
supply line to containment after
a main steam line break.
The
NRC
approved
a relief request to allow the licensee to disassemble
and
inspect the valves
on
a rotating refueling outage
frequency.
The
licensee
has
added this testing
and inspection requirement to the
IST program.
This item is therefore closed.
Unit I Emer enc
Diesel Generator
EDG l-l Failure to Start
S ecial
Re ort 50-275 89-01
Revisions
1 and
2
0 en Item 89-Ol-XO
Closed
On February I, 1989, during
a routine surveillance
procedure,
l-l failed to start.
Two of the four air start motors failed as
a
result of overload of the pinion retainers,
pinion retainer bolting
and rotor shafts.
The licensee
entered
TS action statements
corresponding
to an inoperable diesel generator,
and
commenced
investigation
and repair.
The licensee attributed the cause of the
failure to inadequate fit-up of the pinion gear
and rotor shaft
during plant maintenance,
resulting in improper torque transmittal.
The licensee identified several
possible contributory causes,
including excessive
Train "A" starting air pressure,
improper vendor
-12-
tolerances for slot fabrication of the pinion retainer,
inadequate
vendor assembly instructions for the motor,
and
an inadequate
pinion
retainer bolt locking mechanism.
Revision
1 of this special
report
provided inspection results for the remaining
18 air start motors,
which identified
11 additional air motors with incorrect fit-up and
10 with rotor cracks.
The licensee
documented corrective action for each of the above
listed causes,
and performed engineering
analysis of the design of
the diesel air start configugation.
The licensee
concluded that the
air start
system
was adequate
and, in conjunction with 'corrective
actions,
was acceptable.
The licensee
discussed
their findings with the vendor.
The vendor
indicated that they were not aware of any similar failures.
The
licensee
informed other licensees
of their findings through
an
industry problem communications
system.
This issue
was also the
subject of NRC Information Notice 89-84.
The above actions
appeared
appropriate
This item is closed.
'Unit
1 Fuel Handlin
Buildin Ventilation I o crab
e
Ou in
Fue
Movement
LER 50-275 89-
9
Revision
1
Closed
On, January
18,
1991, the Unit
1 fuel handling building failed to
meet the negative I/8-inch water gage pressure
acceptance
criterion.
The licensee
determined that the fuel handling building ventilation
system
had
been inoperable
since October
15,
1989,
when fuel
movement occurred during the Unit
1 third refueling outage.
Investigation of the root cause of failure identified sources of
leakage to be gaps in sheet
metal siding, unsealed
piping
missing or degraded
door seals,
gaps
between
the
movable crane wall and the fixed walls, through-wall oxidation of
sheet
metal siding,
and reduced flow through the ventilation ducts
due to accumulation of debris.
Because
sealing of the pressure
boundary provided adequate
negative pressure,
the root cause
was
considered
to have
been degradation of the building seals.
P
Based
on observation of the implementation of the design
changes
on
the fuel handling building seals,
and licensee
action which has
maintained continuing awareness
of fuel handling building
ventilation operability, this item is closed.
Ino erable Unit
1 Valve Due to Limitor ue
S rin
Pack Relaxation
LER 50-275 90-09
Revision
0
Closed
In March,
1990, during overhaul of a safety injection system valve,
a spring pack for a motor operator
was found to have
a less than
expected preload.
This condition was also identified on two other
valves,
and was verified to be the result of spring pack degra-
dation.
Nine valves in each unit were identified as potentially
affected.
In May 1990, the vendor,
Limitorque, issued
a Maintenance
Update Bulletin which indicated that valve models other than
SMB-0
-13-
may also have spring pack. relaxation.
After review of maintenance
records
and valve diagnostic information, the licensee
determined
that no additional valves were affected at Diablo Canyon.
The
licensee
replaced
each of the potentially affected springpacks,
and
evaluated
the safety significance of the spring packs which had been
identified as degraded.
Based
on the
as found evaluation,
the
licensee
concluded that the affected valves would have performed
their safety function.
Based
on this assessment
and the prompt
corrective action, this item is closed.
Failure to Perform Houri
Un t
LER 50-
5'
-15
Revision
0
Closed
On September
17,
1991,
a roving hourly fire watch was not performed
in some of the safety related
equipment
rooms.
The fire watch was
delayed leaving the radiologically controlled area,
aqd was not able
to contact his supervisor in time to prevent violation of the
Technical Specification requirement.
For corrective actions,
an incident summary
was prepared
and
reviewed with fire watch personnel,
and included in fire watch
training.
Written instructions
have
been provided to personnel for
actions to take if delays occur on rounds.
Therefore this item is
closed.
Ino erable Unit
1
EDGs While Fuel
Pool
Crane
Was 0 e at
Over
uel
Pool
LER 50-275 91-14
Revision
0
Closed
On February
13,
1991, with the core offloaded, all three
became
During this time, irradiated fuel was moved in the
spent fuel pool.
Although
a specific operational
mode was not in
force, the licensee
determined that the intent of TS 3.8.5.2.b.
(Electrical
Power Requirements)
was not met, since loss of offsite
power would have resulted in loss of fuel handling building venti-
lation.
The licensee
returned
EDG 1-2 to service
on February
13.
For corrective action, the licensee
issued
procedural
guidance to
interpret electrical
power requirements
to include operational
conditions
when the core is offloaded.
This corrective action
appeared
appropriate.
This item is closed.
Hissed
Rod Position Indication Surveillance
Unit
1
LER 50-275
91-17
Revision
0
Closed
From November
14 to November
16,
1991,
rod position surveillance
tests
were missed
as
a result of inadequacies
in the plant process
computer
(PPC) software.
After the
PPC was shut
down and restarted
on November
14, the
PPC incorrectly updated
the reactor trip
breakers'tatus
as open.
Therefore,
the logic in the
PPC inhibited
the alarm function of the rod position deviation monitor.
Licensee
corrective actions included assessment
that rod position had
been
correct during the interval of missed surveillance,
revision of the
PPC software,
operator training to identify rod position deviation
monitor inoperability,
and revision of procedure
AP C-27,
Computer
Software guality Assurance,
to ensure significant computer functions
-14-
are verified on an appropriate
schedule.
These actions
appeared
appropriate,
and this item is closed.
Unit I Low Vacuum Turbine Tri
and Subse
uent Reactor r'ue to a
Pro rammatic Deficienc
LER 50-275 .92-04
evision
0
0 en
This event is described
in paragraph
4.a of this report.
The
LER
will be left open for followup of licensee'orrective
actions
related to posting local procedures.
Ino erable Unit 2 Fuel Handlin
Buildin Ventilation
S stem
urin
Fuel
Movement
LER 50-323 90-02
evision
2
Closed
As a result of inadequate
planning
and understanding
of fuel
handling building ventilation system operability requirements,
the
licensee
moved fuel in the spent fuel pool while the fuel handling
ventilation boundary doors were blocked
open with temporary
hoses
used for outage work.
As
a result, the required negative pressure
was not adequately
maintained.
The licensee
reviewed this event
with Operations
and Maintenance
personnel,
and posted
signs
on all
affected doors
announcing
the need to ensure that doors are closed
during fuel movement.
This corrective action appeared
adequate;
therefore, this item is closed.
Corrective action for the generic
concern of personnel
errors will
be addressed
in Followup Item 50-275/91-27-02,
documented later in
this report.
This item is therefore closed.
Two Unit 2 Main Steam Safet
Valves Set With Ino erable
est
E ui ment
LER 50-323 91-02
Revi sip
0
C osed
On August 26,
1991, the licensee
performed routine verification of
main steam safety valve setpoints.
After two valves were adjusted,
it became
apparent that the test equipment
had failed.
First,
had been adjusted
+3 flats, roughly a 33 pounds per square
inch
(PSI)
change in setpoint.
Then,
RV-225 had been adjusted
13 flats,
roughly corresponding
to 143 psi setpoint
change.
Because of the
unexpectedly
high magnitude of adjustment,
the licensee
suspected
and promptly confirmed improper performance of the test equipment.
Valve RV-225 was declared
However, the licensee
did not
declare valve RV-60 inoperable,
based
on the small,
expected
magni-
tude of adjustment.
Subsequent
testing
found that RV-60 lifted
4 psi
above its +/- 1X tolerance
range.
The licensee
concluded that the test equipment
had failed abruptly
while setting
RV-225,
and
had
been operable while setting
RV-60.
The licensee
also concluded that there
was
no firm evidence that the
test equipment
caused this out-of-tolerance setpoint for RV-60,
since setpoint drift is
common
and expected
up to +/-3X.
Because
RV-60 was not declared
inoperable after test equipment failure was
confirmed, the
NRC concluded that this was
a non-cited violation
(50-323/91-24-02),
based
on the low safety significance of the
as
found setpoint of RV-60, the licensee's
prompt corrective action to
fix the test equipment
and validate the actual safety valve set-
-15-
points,
and continuing improvement in documentation of the rationale
for operability decisions.
Therefore, this item is closed.
Inadvertent
Removal of Power
From Unit 2
Pum
Sum
Suction Valve
0 erators
Before Enterin
Mode
5
LER 50-323 91-03
ev sion
0
Cl osed
On September
1,
1991, control
room operators
determined that both
and both containment
spray
(CS)
pumps were
The
RHR valve breakers
had
been
opened locally, and
control power to the
CS pump breakers
had been
removed;
Both
events'ccurred
after entry into hot shutdown earlier in the day,
and were
corrected within 15 minutes after identification by operators.
For the
RHR valves,
power must
be removed from the 'Valves at the
control
room contactor toggle switches
(rather than the local
breakers)
to satisfy safety analysis
requirements
that power be able
to be restored
from the control
room.
Procedure
OP L-5 required
power be removed from the valves at the local breaker,
thus
violating safety evaluation requirements.
The licensee
revised
OP
L-5 and reviewed all other L-series procedures.
For the Containment
Spray pumps,
power was removed before entry into
Mode
5 as
a result of inadequate
control
and review of clearances
by
a Senior Control Operator.
Specific licensee corrective actions
included issuing
an Operations
Incident Summary
and memoranda
stressing
the significance of this
event
and the need for ensuring
TS commitments
are met.
Corrective
action for the generic concern of personnel
errors will be addressed
in Followup Item 50-275/91-27-02,
documented later in this report.
This item is therefore closed.
Shift of Unit 2 Control
Room
Ve tilation Mode
LER 50-323 91-06
Revision
0
Closed
On October 3,
1991,
an inadvertent
engineered
safety features
(ESF)
actuation occurred.
This was the result of inattention to detail
by
two operators
who lifted an incorrect lead while attempting to
remove
an inverter from service.
Power was interrupted to inverter
IY-23, resulting in a shift of the control
room ventilation system
to the pressurization
mode.
All safety systems
functioned
as
expected;
The operators
followed generic procedure
guidance
instead of
detailed information provided in an attachment.
As corrective
action,
both operators
were counseled
regarding attention to detail,
and the procedure
was revised to delete the generic steps
and
emphasize
detailed information.
Corrective action for the generic
concern of personnel
errors will be addressed
in Followup Item
50-275/91-27-02,
documented later in this report.
This item is
therefore closed.
0
-16-
Inadvertent Unit 2 Safet
In'ection While in Mode
5
LER 50-323 91-
07
Revision
0
Closed
On October 6,
1991, during a refueling outage
(Mode 5), technicians
inadvertently caused
a safety injection signal
by performing
protection switch positioning out of the correct
sequence.
The
cause of the event
was inattention to detail.
All expected
actuation
signals occurred.
No water was injected into the core,
since
emergency
core cooling pumps were secured for maintenance.
The technicians
were counsel1ed
according to the licensee's
positive.
discipline program,
and
a memorandum
was is'sued
by the Plant Manager
emphasizing
the need for procedure
compliance
and verification.
Corrective action for the generic
concern of personnel
errors will
be addressed
in Followup Item 50-275/91-27-02,
documented later in
this report.
This item is therefore closed.
Unit 2 Valve Leaka
e
Ma
Have Resulted
in Exceedin
10 CFR 100 Doses
LER 50-323 91-09
Revisions
0 and
1
Closed
On October 4,
1991,
1.3 gallons per minute leakage
from diaphragm
valves
CVCS-2-8471
and -548 was evaluated
by the licensee.
The
evaluation
concluded that the control
room and exclusion area
boundary
10 CFR 100 limits could have
been
exceeded
during the
recirculation
phase of a loss of coolant accident
(LOCA).
The licensee
determined that the diaphram service life for CVCS-2-
8471
had
been
exceeded
and that diaphram degradation
had resulted in
leakage.
The
LOCA analysis
had not identified CVCS-2-8471
as
a
valve in the recirculation
phase flow path.
As a result,
the valve
was not included in the licensee's
preventive maintenance
program
for diaphragm replacement.
The licensee
determined that the body-to-bonnet bolts for CVCS-2-548
had not been retorqued at normal operating pressure
and temperature
following maintenance,
as
was required for that type valve.
The
licensee
committed to revise the procedure to include this step.
While this specific issue is closed,
the inspector
was concerned
that other issues
concerning the design requirements
of the
recirculation
phase
may not have
been adequately
addressed,
and
documented
the need for
NRC followup of this concern earlier in this
report.
Ino erable Unit
1 Wide Ran
e Containment
Sum
Level Indicatio
LER 50-323 91-10
Revisions
0 and
1
C osed
On October 22,
1991, with Unit 2 in Mode
2 at
OX power,
TS Action
Statement
3.3.3.6
was exceeded
due to the reactor cavity sump wide
range level channel
being inoperable for more than
seven
days.
The
channel
was returned to service
about
7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> later.
-17-
Failure to met the
TS action statement
was caused
by inadequate
corrective action for a previous similar occurrence.
During investigation of a similar failure on March 15,
1992,
the
licensee
determined that only the zero range
was affected
by the
failure,
and all the ranges
required for the indicator to fulfill
its safety function were operable.
The licensee
considers
previous
failures to be similarly of low safety significance.
In Inspection
Report 50-323/$ 2-01, the
NRC issued
a violation
concerning the inoperable
sump level indication.
This 'item is
closed.
Failure to Test
a Unit 2
ECCS Valve After Maintenance
LER 50-323
91-11
Revision
0
Closed
On October
17,
1991, while in Mode 5, the packing
on
ECCS valve
SI 2-8802B was adjusted
and the packing gland follower nuts were
torqued to a value greater
than the .torque
used for the last
successful
stroke time test.
Both these
operations result in
requirements
to perform a post-maintenance
test before returning the
valve to operable status.
However, neither requirement
was
immediately recognized.
During mode change
review on October 22,
the deficiency was recognized,
and the test
was performed.
As corrective action, training on mode transition
was provided for
the Engineering Test Group,
an Operations
event report was issued,
and work planning issued
enhanced
work policies.
Corrective action
for the generic
concern of personnel
errors will be addressed
by
Followup Item 50-275/91-27-02,
documented later in this report.
This item is therefore closed.
11.
0 en
a ~
Item Followu
92703
Public Address
S stem
Powered
b
Non-Vital Circuitr
Followu
Item 50-275 91-09-03
Closed
b.
During the loss of offsite power event
on March 7,
1991, operators
were unable to quickly inform workers in the containment of plant
status
because
the
PA system
was powered from non-vital power.
Containment
was evacuated
in an orderly manner without use of the
system.
The licensee
plans to implement control of PA system
power
sources
during outages.
Based
on planned licensee
actions, this
item is closed.
EDG Slow to Load
Followu
Item 50-275 91-09-04
C osed
This item discussed
the failure of EDG l-l to load onto the vital
bus within 10 seconds
during the loss of offsite power event
on
March 7,
1991.
After NRC review of the circumstances,
the licensee
received
a violation associated
with determination of the validity
of the start failure (50-275/91-07-10).
No other issues
were
identified.
Therefore, this item is closed.
0
18
Refuel in
Procedures
Chan
e
Foll owu
Item 50-275 91-09-05
Closed
During the loss of offsite power on March 7,
1991,
a concern
was
identified regarding the safety of fuel assembles
in transit during
The licensee identified a commitment
based
on
INPO SOER 85-01 to change
procedures
to include actions to place
fuel assemblies
in a safe, location during loss of refueling cavity
cooling concurrent with a loss of power.
The inspector reviewed Revision ll of procedure
OP B-BDS2, Core
Loading Sequence,
which was %evised to include
a step which does
not.
allow a fuel assembly to be lifted from the spent fuel pool until
the previous
assembly
has
been unlatched in the core.
The inspector verified that
a similar requirement
eMisted in the
core unloading procedure.
Based
on the licensee's
corrective
action, this item is closed.
Ino erable
Wide Ran
e Reactor Cavit
Sum
Followu
Item 50-323 92-
01-03
Closed
The
NRC issued
a Notice of Violation due to inoperable
wide range
reactor cavity sump level indication.
Since that time, the licensee
has successfully identified and corrected failure of the
sump level
indication
on two occasions,
as
a result of more attention to
Technical Specification operability requirements.
As permanent
corrective action,
continued troubleshooting
and repair work on the
level indication has occurred during unscheduled
outages.
A design
change
has
been reviewed for installation during the next scheduled
outage.
Based
on the licensee's
corrective actions to date,
and
planned corrective action, this item is closed.
Increase
in Personnel
Errors
Followu
Item 50-275 91-27-02
0
e
In addition to followup of existing licensee efforts to reduce the
number of personnel
errors, follow of corrective actions for this
item will include the generic corrective actions for licensee
event
reports
documented
above which were caused
by human error.
Indica-
tion of successful
actions will include reduction of the number of
LERs caused
by human error.
This item remains
open.
S urious Actuation of Carbon Dioxide Fire
Su
ressio
S stem in
1-3
Room
Unreso1ved
Item 50-275 92-05-01
Closed
During a routine monthly surveillance of the 1-3
dioxide (cardox) suppression
system actuated.
The ventilation for
generator
cooling was then cut off as
a result of the cardox system
closing the west rolldown doors.
Operators
shut
down the diesel to
prevent overheating.
The
NRC questioned
whether this was
a valid
failure of the diesel.
Because
the root cause is still under
investigation,
and because
the licensee is reviewing the validity of
failure evaluations
in response
to an
NRC notice of violation, this
issue is closed.
Followup of the issue of validity of failure will
be part of routine resident
inspector activities.
-19-
g,
Com arison of Li uid Radwaste
Anal sis Results
Followu
Item
50-275 88-33-01
Closed
This item concerned
an inter-comparison of i-ron-55 (Fe-55) activity
in a liquid radwaste
sample that was split between
the licensee
and
the
NRC contract laboratory.
The initial inter-comparison of
analytical results for Fe-55 did not agree,
and
a followup
inspection
was conducted
at the licensee's'off-site
laboratory.
The
inspector found that the licensee's
laboratory sampling
and measure-
ment techniques
were fundamentally
sound.
To resolve the issue,
the
inspector,
in coordination with the licensee,
devised
an experiment
'o
independently verify both NRC's
and the licensee's
analytical
methods for determining
Fe-55 in liquid radwaste.
A liquid radwaste
sample,
spiked with FE-55,
was split between the
NRC contract
laboratory
and the licensee.
The amount of FE-55 activity was
known
only by the inspector.
Based
on the results of the inter-comparisons,
the inspector
concluded that the licensee's
method for analyzing liquid radwaste
for FE-55 is satisfactory.
This item is closed.
12.
Followu
on Unresolved
Items
92701
a ~
uxiliar Salt Water
ASW
Floodin
Due to Failure to Follow Proce-
dures
U
eso
ved Item 50-275
9 -24-01
C osed
On August 2,
1991, flooding of the component cooling water heat
exchanger
area occurred
as
a result of maintenance
worker s'ailure
to follow clearance
instructions in the order specified.
Review of
the safety significance of this issue determined that the other
trains of CCW and
ASW were available
and would have performed
their safety functions.
Licensee corrective actions
included issuing
a maintenance
event
summary,
counseling the individuals involved using positive
counseling,
and emphasizing
the personal
and plant safety aspects
of
properly following clearances.
The failure to follow clearance
instructions
appears
to be
a viola-
tion of Technical Specification 6.8. 1, which requires that work be
implemented
by procedures.
Based
on the licensee's
corrective
action
and the low safety significance of the concern,
the violation
is not being cited because
the criteria specified in Section V.A. of
the Enforcement Policy were satisfied
(92-16-05, closed).
Se aration of Safet
Related Electrical Circuits
Unresolved
Item
50-275 91-11-01
Closed
The licensee
had previously identified that two independent
Class
1E
direct current (dc) supplies
were electrically connected
in a non-
Class lE fire horn relay panel.
There
was
no electrical
separation
within the panel.
0
-20-
The licensee
concluded that connection of the two Class
1E dc
supplies in the relay panel
was within the separation criteria
established
by their license.
r
An inspector also previously noted that both Class
1E cables
and
non-Class
1E cables
were being routed through
a new Conax penetra-
tion.
The licensee
stated that they were using Regulatory Guide (RG) 1.75,
"Physical
Independence
of Elect'rical Systems,"
when
feasible during design modifications.
The licensee
stated that they
were using coax cables, for phich there were
no spare penetrations.
Therefore,
the licensee
concluded that separation
per
RG 1.75 was
not feasible.
The inspector determined that the routing of both Class
lE cables
.
and non-Class
1E cables
through the
new Conax penetration
would be
further reviewed for acceptable electrical separation.
The review
would be performed in conjunction with the review of the accepta-
bility of electrical separation within the fire horn relay panel.
The inspector evaluated
both of the previously identified items
discussed
above.
The inspector noted that the Diablo Canyon design
predated
The'esign
basis for Diablo Canyon included
IEEE 308-1971, "Criteria for Class
lE Electrical
Systems
for Nuclear
Power Stations."
Paragraph
5.3.5 of IEEE 308-1971 stated that
protective devices shall
be provided to isolate failed equipment
automatically.
The inspector reviewed the circuit design
and
concluded that the Diablo Canyon design for protective devices
was
adequate
to prevent faults in non-lE Class
panels
from affecting
Class
1E dc supplies.
The inspector also concluded that the installation of both Class
1E
and non-Class
1E cables in a new Conax penetration
met the design
criteria for Diablo Canyon.
The inspector
concluded that lack of a
spare penetration
was
a reasonable
evaluation for concluding that
installing the
new cables
per
RG 1.75 criteria was not feasible.
This item is closed.
4 Kilovolt Switch ear Fault Current Ratin
Unresolved
Item 50-275
and 50-323 91-07-01
Closed
The Electrical Distribution System Functional
Inspection identified
that calculated fault current exceeded
4 kilovolt (kV) switchgear
ratings during certain plant operations.
Calculated fault current
exceeded
switchgear ratings
when one or more emergency diesel
generators
(EDGs) were operated
in parallel with the main generator.
The licensee
took action to minimize the tests
which required para-
llel operation of the main generator
and one or more
EDGs.
The
licensee
also took action to minimize the fault capability of the
main generator
during parallel operation with an
EDG.
-21-
The licensee
performed
a calculation which showed that
a maximum
(bolted) fault was
a low probability event during the limited time
the main generator
was operated
in parallel with an
EDG.
The
NRC staff calculated the core
damage
frequency
caused
by the
occurrence of a bolted fault during
a time when the main generator
was operated
in parallel with an
EDG.
The calculation
showed that
the core
damage
frequency
caused
by this bolted fault was in the
order of 1.0E-9 per reactor year or less.
The inspector reviewed the 1'Icensee's
actions
and concluded that
these actions to minimize the risk from a bolted fault were
adequate.
The inspector reviewed the licensee's
calculation
and
concluded that this calculation
was adequate
to demonstrate
that
a
bolted fault during parallel operation of the main generator
and
an
EDG was
a low probability event.
The inspector also concluded that
the core
damage
frequency calculation demonstrated
that the bolted
fault condition discussed
above
had little safety significance.
This item is closed.
13.
Verification of As-Built Drawin s
37051
and
50073
On April 24-25,
1992, the inspector entered
containment to examine
containment
fan cooler damper 2-4 to verify the licensee
had restored
this component to the requirements of the design drawings
and documents.
The inspector
used the following documents
as the basis of this
inspection.
~
American Warming 5 Ventilating, Inc., Drawings SHW-D-9098,
PGKE
Revision 6, record 663079,
Sheet
37;
SHW-D-9099, Revision
B; and
81010-001-000
Revision A,
PGKE record 663079,
Sheet
75
~
Design
Change Notice DC2-EH-44664,
CFCU Backdraft Damper Helper
Springs
~
Non-Conformance
Report DC0-92-HH-N007, Containment
Fan Coolers
The inspector
observed that the backdraft
appeared
to conform to
the design
documents with two exceptions.
First, several of the Allen
head shoulder bolts holding the horizontal linkage
arm to the individual
vane linkage
arms
appeared
loose in the holes in the linkage arms.
Second,
about half of the counterweights
on the ends of the linkage arms
did not contact the rubber block shock absorbers
when the damper
was
closed.
The gap
was 1/16 inch or less.
The design
documents
did not
specify any clearances
related to these observations.
When questioned
by the inspector,
the Hechanical
Maintenance
Engineer
showed the inspector
an Action Request
which had
been written regarding
the loose fit of the shoulder bolts in the linkage arms.
The licensee
concluded that this deficiency would not have prevented
the damper from
working and planned to completely overhaul
these
during the next
refueling outage.
-22-
The inspector re-entered
containment to observe the surveillance testing
(M-51A) of damper 2-4 to determine if the lack of contact with the shock
absorber
block would affect the operability of the damper.
The inspector
observed that starting
and stopping the associated
containment
fan cooler
unit caused
the damper to swing open
and .shut over
a period of several
seconds.
The inspector concluded that during normal operation the damper
would not be significantly affected
by the counterweights
not resting
on
the shock absorber
block.
The inspector requested
that 'the licensee
evaluate
the effect of the observed
gap
on the post-LOCA performance of
the dampers.
The licensee
evaluated
the observation
and concl'uded that during
a
LOCA,
with the gaps
measured
by the licensee,
the flexibilityof the linkage
mechanism
and the dampers
was sufficient so that the shock absorbers
would absorb
any significant shock to the linkage arms.
'o
violations or deviations
were identified.
14.
nformation Meetin
with Local Officials
94600
On May 4,
1992, the resident
inspectors
met with the
San Luis Obispo
County Administrator, Deputy Administrator,
and
Emergency
Response
Coordinator.
The purpose of the meeting
was to introduce the
NRC
inspectors,
to offer the NRC's openness
to discuss
inspection report
findings with the County,
and to discuss
general
issues of emergency
~
~
~
~
~
planning.
15.
~Fit
M
An exit meeting
was conducted
on June
12,
1992, with the licensee repre-
sentatives
identified in Paragraph
1.
The inspectors
summarized
the
scope
and findings of the inspection
as described
in this report.