ML17333A585
| ML17333A585 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/27/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17333A583 | List: |
| References | |
| 50-315-96-10, 50-316-96-10, NUDOCS 9610030336 | |
| Download: ML17333A585 (15) | |
See also: IR 05000315/1996010
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION III
Docket Nos:
Licenses
No:
50-315,
50-316
Reports
No:
50-315/96010(DRS);
50-316/96010(DRS)
Licensee:
Company
Facility:
Donald
C.
Cook Nuclear Generating
Plant
Location:
7700
Red Arrow Highway
Stevensville,
MI
49127
Dates:
August 5-29,
1996
Inspectors:
Approved by:
D. Butler, Reactor Inspector, RIII
R. Lerch, Reactor Inspector,
RIII
M. A. Ring, Chief
Lead Engineers
Branch
Division of Reactor Safety
Ins ection
Summar
Routine inspection of previously identified issues
from the Integrated
Performance
Assessment
Process
and other inspections.
One violation for lack
of test control
was identified for lack of surveillance testing of some
component cooling system flows.
One unresolved
item was initiated regarding
the appropriateness
of a revision to the Updated Final Safety Analysis Report.
96f0030336
960927
ADOCK 050003i5
8
O
M8.1
Miscellaneous
Maintenance
Issues
Closed
Violation 50-315 95014-01
The licensee
did not properly
train Instrument
and Control (IC) technicians
and provide adequate
calibration details in maintenance
procedure
No.
12
IHP 6030
"Protective Relay Calibration."
In response,
the licensee requalified all of the
IC technicians that
calibrate protective relays.
In addition, the maintenance
procedure
was
revised with a note defining the relay pick-up current
as the point
where the continuity light "just flickers."
The inspectors
reviewed the
technician training records,
training lesson
plan No. IC-C-0720,
"Protective Relays," the maintenance
procedure,
and observed
two.
protective relay calibrations
in the field.
The inspectors
concluded
the licensee
had addressed
this violation in an acceptable
manner.
In
addit1on,
the
IC technicians
that performed the calibrations clearly
understood
proper protective relay calibration techniques.
This item is
considered
closed.
E2
E2.1
Engineering
Support of Facilities
and Equipment
Follow-u
of NRC Issues
Re ardin
Com onent Coolin
Water
S stem
Flow Balancin
'a
~
Ins ection
Sco
e
2.
In February
1996,
an Integrated
Performance
Assessment
Process
(IPAP)
inspection
team identified that the completed Unit
1
CCW flow balance
procedure,
1
EHP 4030 STP.248,
"Unit
1 Component
Cooling Water
Flow
Balance,"
completed
on September
28,
1995,
was inadequate
because:
Sample cooler flows were not met,
although objectives
were provided in
the procedure.
The objectives
were not acceptance
criteria, although
the flows were described
as the minimum required flows in the Updated
Final Safety Analysis Report
(UFSAR).
Cooling flow to the containment air recirculation fan was not verified
by the surveillance.
Additional details
on these
observations
may be found in Inspection
Reports
No. 50-315/316-96003
and
No. 50-315/316-96006
(Inspection
Follow-up Item (IFI) 96006-03).
The inspector
reviewed licensee
actions
taken since conclusion of the
IPAP inspection with regard to
surveillances
and discussed
the issues
with the licensee's
staff.
The
inspectors
also reviewed the
UFSAR revision dated July 1996
and
a safety
evaluation
performed regarding
the
CCW system description revision to
the
2
Observations
and Findin
s
A Unit 2 flow balance surveillance,
2
EHP 4030 STP.248,
"CCW Flow
Balance,"
was started
in April and completed
Hay 1,
1996,
using
a
procedure similar to the one reviewed
by the
IPAP team.
No changes
in
response
to
NRC concerns
had
been
made;
however,
the
IPAP inspection
report
had not been received/reviewed
by the licensee.
The
IPAP report
was issued April 17,
1996.
Nonetheless,
inspector interviews with
members of the engineering staff indicated the requirement for the
surveillance
program to verify and maintain
system performance
in
accordance
with the
UFSAR was not well understood
or appreciated
by
several
members of the engineering staff.
A UFSAR revision dated July 6,
1996,
was submitted to the
NRC, which
revised Table 9.5-2,
making miscellaneous
flows (sample coolers,
thermal barriers,
seal injection,
and others) for information only
rather than minimum required flows.
The licensee
completed
a safety evaluation
dated July 31,
1996, to
evaluate
the July
UFSAR revision.
The position of the safety evaluation
was that only safe
shutdown cooling loads
needed
to have
minimum
requirements
specified.
Issues
such
as license
commitments
made to meet
Three Mile Island action items for post-accident
sample cooling, or the
importance of reactor coolant
pump thermal barrier cooling to potential
events
were not,addressed.
Given the importance of miscellaneous
cooling flows to plant operations,
the designation of flow information
as "for information only" could
be misleading.
Discussions
with the licensee
indicated that based
on recent
system flow
readings
and sampling
by the chemistry department,
the sample coolers
were useable
with reduced
coolant flow under normal conditions.
Additional cooling of samples
could
be obtained
by slowing the flow of
the sample
through the cooler.
Conclusions
The inspectors
identified the following issues:
The containment recirculation/hydrogen
skimmer fan cooling flow is not
verified by any surveillance.
The
CCW sample cooler flows were measured
by surveillances
1
EHP 4030
STP.248,
"Unit
1 Component
Cooling Water Flow Balance,"
completed
on
September
28,
1995,
and
2
EHP 4030 STP.248,
"CCW Flow Balance,"
completed
Nay 1,
1996,
but measurements
indicated they were less
than
required
by the test objectives
and the
The chemistry staff
stated that samples
can
be adequately
cooled with existing flows;
however,
the ability to cool
samples
under accident conditions
was
undetermined
and
had not been evaluated.
When engineers
created
surveillance
procedure objectives
and accepted
surveillance
flow results
which were less
than the
UFSAR specified for
0
CCW flow to sample coolers,
they changed
the operation of the system
as
described
in the
UFSAR ~
Although the effects
appear
to have
been
minimal, they were not evaluated
and safety evaluations
were not
performed.
The inspectors
were concerned that it was not appropriate for the
revision to describe
the flow requirements
for the miscellaneous
as nominal
and for information only.
Issues
1 and
2 were examples of inadequate
test control.
The licensee
initiated condition report 96-1367 to track corrective actions for these
issues.
The failures to verify adequate
CCW flows through surveillances
were examples of a violation of 10 CFR 50, Appendix B, Criterion XI,
"Test Control" (50-315/316-96010-01).
Issue
3 was related to the
concerns of Issue
2.
E2.2
Issue
4 required further review by the
NRC to determine
the information
and level of detail
needed
in the
This is an unresolved
item
(URI) pending further review by the
NRC (50-315/316-96010-02).
NRC Information Notice
Potential for Loss of Remote
Shutdown
Ca abilit
Durin
a Control
Room Fire
The
NRC notified the industry via IN 92-18 about
an unanalyzed
condition
regarding fire protection
and
a plant's safe
shutdown capability during
a control
room fire that forced reactor operators
to evacuate
the
control
room.
This fire location could cause
such
as short
circuits between
motor operated
valve
(HOV) control circuit conductors
and their control
power source.
As a result,
spurious operation of
certain
MOVs may occur before the operators
shifted control of the
valves to the remote/alternate
shutdown panel.
Motor thermal
overload
protection
may be bypassed,
set high or set with a longer tripping time
to allow for additional
valve duty cycles and/or reversing of the
HOV
during stroking.
The
IN identified that
HOV torque
and limit switches
=would not electrically disconnect
the stroking valve, causing
mechanical
damage
to the valve and/or damaging
the motor due to the hot short
bypassing
the limit or torque switches.
Licensees
took credit for
manual manipulation of certain
HOVs in many of the Appendix
R safe
shutdown scenarios,
but in some instances,
did not consider the valves
could
be mechanically
damaged.
The licensee
provided the inspectors
a list of electrically operated
Appendix
R valves.
The D.
C. Cook'design
uses
a double break
scheme
around the valve open/close
In most cases
this was provided
by both limit/torque switch contacts
and remote/alternate
transfer
switch contacts.
For the above,
two hot shorts of the correct polarity
and conductors
must occur to operate
a valve.
The inspectors
reviewed
the electrical
design of the Appendix
R valves for both units.
All the
valves incorporated
the double break design.
The inspectors
noted that
four steam generator
(SG) turbine driven auxiliary feedwater
pump
(TDAFP) feed valves
had
common relay contacts,
such
as relay 63Xl-SGlT
for the valve feeding
SG1, that could initiate valve operation
due to
one hot short.
However, the torque switch remained electrically
connected
and would prevent mechanical
damage to the valve.
The
emergency
remote
shutdown
procedure
had the operators electrically
disable the individual
TDAFP to
SG valve power supplies.
This would
effectively remove
any continuous
close signal
and permit manual
valve
manipulation.
In addition, only two SGs were required for an Appendix
R
scenario.
The probability of a single hot short
on more than
two TDAFP
feed valves
was extremely low.
The inspectors
concluded that
D.
C.
Cook
met the intent of the IN.
Susce tibilit of Ta lor MOD 30 Di ital Controllers to Electrostatic
Dischar
e
On March 17,
1996, the differential pressure controller for the Unit I
feedpump
(I-RU-05) failed without operator interaction
and resulted
in a
Unit conditions were stabilized
and the controller was
replaced.
Initial investigations identified that the controller had
failed in the unconfigured state.
During this period,
several
other
controllers exhibited control
panel "flicker" when touched.
The
licensee's
root cause
analysis
determined that
ESD was causing
the
30 controller problems.
The licensee
concluded that controller circuit board static drain clips
were not properly oriented
and were not grounding the controller.
In
addition, the control
room humidity was very low and
had increased
buildup.
The licensee
had installed
ESD limiting carpeting.
Samples
were sent to the manufacturer to verify that the correct carpet
had
been
installed.
The test results
indicated that the installed carpet
was
limiting (l. I KV, step
and 0.9 KV, scuff).
In addition, the licensee
installed grounded static step-off pads in front of ESD sensitive
equipment.
The
MOD 30 manufacturer
performed
ESD sensitivity tests
on the failed
controller.
ESD levels to
3
KV were capable of inducing control
panel
"flicker."
The manufacturer installed
an additional
ground wire from
the module mounting case that increased
ESD tolerance to
7 KV.
The
extra ground wire has
been installed
on the feedpump controller at
D.
C.
Cook.
In addition, the manufacturer
has identified that the electrical
conductor ribbon strip that attached
to the membrane
keypad switch
control
panel
(face-plate)
was also susceptible
to
ESD.
The
manufacturer
was testing different face-plate modifications.
The
immunity has
been
increased
to
12
KV by insulating the ribbon strip.
Replacement
of the currently installed face-plates
was being reviewed
by
the licensee.
The licensee
has initiated other actions to improve
ESD immunity.
These
include control
room humidity monitoring
and improving control
room
humidifier operation.
The inspectors
concluded that the licensee
was
effectively addressing
ESD affects
on digital equipment.
Niscellaneous
Engineering
Issues
Closed
IFI 50-315 316-95005-05
During the past several
years,
numerous
N-2AO-V2H (V2H) voltage to current cards
have failed
due to design
and/or component related
weaknesses.
These
included
an
undersized circuit board fuse
and the encapsulated
(potted)
DC to
converter
module.
The licensee
has worked closely with Foxboro to
resolve these
issues.
The resultant
design
changes
have replaced
the
fuse with a larger size
and
added additional
fuse protection.
The
DC to
DC converter
module
was completely redesigned
and installed
on all of
the safety related cards.
Only one
new card failure has occurred.
This
was attributed to
a card burn-in type failure.
The licensee
indicated that the
new
DC to
DC converter
module
had
a
manufacturing flaw.
Two surface
mount biasing resistors
were installed
at
a lower resistance
value than called for in the design.
Testing
has
demonstrated
that the lower resistor values
have slightly increased
the
module heat generation.
However,
module components
thermal
design
margin was high (= 50X) so that this flaw should not affect
functionality.
The licensee
indicated that card life testing will
continue to determine if the
mean time before failure of the cards
had
increased.
This item is considered
closed.
Closed
IFI 315 316-93012-Ola
and
c DRS:
The Systems
Based
Instrument
and Control Inspection
(SBICI) team determined that certain Westinghouse
"Setpoint Methodology for Protection
Systems
D. C. Cook,
Unit 1," and
"Setpoint Methodology for Protective Systems-
D.
C. Cook, Unit 2," protective action setpoints
did not include
an
environmental
allowance
(EA) term in diverse
secondary
(backup) trip
functions.
The trip functions of interest
included
SG water low level,
low-low level
and high-high level;
and main steam flow/feedwater flow
mismatch.
In response,
Region III requested
assistance
from the Office of Nuclear
Reactor Regulation
(NRR) to determine if backup trip functions were
required to include
an
EA term in the setpoint methodology.
reviewed the
above
backup trips and determined that the accident
analysis did not include
an
EA in the uncertainty calculations.
concluded that the licensee's
methodology concerning
EAs was consistent
with Westinghouse's
methodology
and
IEEE Standard
279; "Criteria for
Protective
Systems for Nuclear
Power Generating Stations."
This item is
considered
closed.
Closed
IF I 315 316-93012-Old
DRS:
The SBICI team noted that the
D.
C.
Cook design
used containment
pressure
to detect
a steam line break
inside containment rather than
steam line flow.
The team was concerned
that this design
was not meeting the intent of IEEE Standard
279-1968,
since containment
pressure
was not
a direct variable measurement
of this
accident condition,
In response,
Region III requested
assistance
from NRR to determine if
containment
pressure
was
an acceptable
variable to detect
a steam line
E8.4
break inside containment.
NRR reviewed
UFSAR Section 14.3.4,
"Containment Integrity Analysis,"
and concluded that the containment
pressure
variable
was the accepted
licensing basis for this accident
scenario.
This item is considered
closed.
Closed
IFI 315 316-93012-Ole
DRS:
The SBICI team noted that
Regulatory
Guide 1.97,
Category
1, condensate
storage
tank
(CST) level
instrument
loop accuracy calculation did not include
a seismic bias
error term.
In response,
the licensee
provided the Foxboro equipment qualification
reports for Model
N-E13DM and
E13DM pressure
transmitters.
The
qualification report
showed that the differential pressure
transmitters
would return within their. reference
accuracy following a seismic event.
NRR concluded that the licensee's
treatment of the
EAs for the
CST level
instrument
loop calculation
was acceptable.
This item is considered
closed.
E8.5
Closed
IFI 315 316-93012-Olf DRS:
The SBICI team noted that the
refueling water storage
tank
(RWST) level instrumentation
cables
were
routed through
(HELB) area.
The D. C.
Cook
UFSAR indicated that the
RWST level channels
were
used to mitigate
an
blow down event.
As such,
an
EA term should
be included for the cable
routed through the
HELB area.
In response,
the licensee
provided
NRR a calculation that included cable
insulation losses.
NRR reviewed the calculation
and concluded that the
licensee
had addressed
cable insulation losses
in an acceptable
manner.
This item is considered
closed.
E8.6
EB. 7
Closed
IFI 50-315 316-96006-03:
CCW flow balance
surveillances
did
not meet the
This is'losed
to violation 50-315/316-96010-01
in
this report.
Closed
URI 50-315 96006-15 DRS:
NRC Inspection
Report
No. 50-
315/316-95010,
Section 3.5, identified that the Unit
1 west motor-driven
pump
(MDAFP) had
a history of instantaneous
relay
(PJC type) trips during Mode 4,
5 or 6 testing.
The testing took place
prior to entering
Mode 3.
The
MDAFP was required to be operable
in
Modes
1,
2 and 3.
In response,
the licensee
issued condition report
No. 95-1204 to address
this concern.
The licensee's
investigation identified that the
MDAFP
,instantaneous
relay (1-50-50N-TA2)
had insufficient margin for motor
starts during high voltage conditions
(4280 volts).
The safety
buses
were supplied
from the reserve auxiliary transformer
(RAT) during Modes
4, 5, or 6 testing.
Since Unit
1 was shut down,
house
loads
were at
a
minimum (< 2.5
MW) while operating plant loads
were typically 20.5 to
36
MW.
This resulted
in higher
bus voltage
due to
a lightly loaded
RAT.
Normal
Mode 1,
2 or 3 electrical
supply to the safety
buses
was from the
unit auxiliary tr'ansformers
that also
power the house loads.
During the
past year's
Mode 1,
2 or 3 operation,
the Unit
1 safety
bus high average
voltage
was about
4195 volts.
The Unit
1 west
pump has tripped four
times since
1987.
All of the trips occurred during Hode
5 or 6 testing
of the pump.
There
have
been over
a hundred successful
Unit
1 west
pump
starts
since
1987.
In addition, the Unit
1 east
and both Unit 2 MDAFPs
were successfully
started
during bus high voltage conditions.
The licensee
determined that the Unit
1 west
HDAFP current transformer
(CT) saturated
at
a higher value than the manufacturer's
saturation
curve.
Therefore,
the resultant
CT excitation voltage
change
was
greater for a given
CT current
change at the
PJC relay.
At higher motor
starting voltages,
an induction motor starting current will increase,
since electrically the motor appears
as
an
impedance
device.
This
generated
higher current flow to the
PJC relay and caused
the relay to
trip sooner.
The four HDAFP PJC relays
were set for a nominal primary current of 27
amperes.
Calculation
PS-4KV-001,
"4 KV Safety Hotor Electrical
Protection,"
developed
the setpoints
based
on
recommendations.
Although the setpoint margin was lowest for the
HDAFP
motors,
the results
were within the calculation acceptance
criteria.
The licensee
increased
the Unit
1 west
HDAFP setpoint to
a primary
current of 33 amperes
and
was reviewing the other
HDAFP setpoints.
In
addition, the licensee
indicated that additional
CT testing
was being
performed.
The licensee
revi'ewed all of the current
4
KV motor
protective setpoints
and concluded the calculation
had
been
performed
satisfactorily.
The inspectors
reviewed the
above information and
CT design standards,
and concluded that the licensee's
root cause determination
and
corrective actions
were reasonable.
In addition, the inspectors
concluded that the Unit
1 west
HDAFP was operable,
since safety
bus
voltage did not exceed
4280 volts during past
Mode 1,
2 or 3 operation.
This item is considered
closed.
V. Nana
ement Neetin
s
Exit Meeting
Summary
The inspectors
met with various licensee
representatives
on August 9,
1996,
and
summarized
the scope
and findings of the inspection
in the
electrical
and instrumentation
and control areas.
Inspectors
also
conducted
an exit meeting
by phone
on August 29,
1996,
and
summarized
the scope
and findings with regard to follow-up on component cooling
water flow balancing.'he
persons listed under the partial list of
persons
contacted
participated
in the telephone exit.
The licensee
did
not identify any of the documents
reviewed
by the
NRC as proprietary.
PARTIAL LIST OF
PERSONS
CONTACTED
Licensee
A. Blind, Site Vice President
K. Baker, Assistant
Plant Hanager
P. Schoepf,
Supervisor,
Safety-Related
Systems
J.
Kobyra, Hanager Nuclear Engineering
R. Ptacek,
Licensing
W. HcCrory,
System
Engineer
INSPECTION
PROCEDURES
USED
On-site Engineering
Haintenance
Observation
ITEHS OPENED,
CLOSED,
AND DISCUSSED
~0ened
50-315/316-96010-01
50-315/316-96010-02
Closed
Lack of test control for CCW flow balance
requirements.
Revision to the
FSAR changed
the minimum
required
CCW miscellaneous
flows to nominal
flows for information only.
50-315/95014-01
50-315/316-96005-05
50-315/316-96006-03
50-315/96006-15
IFI
IFI
The licensee
did not properly train
technicians
and provide adequate
calibration
details.
During the past several
years,
numerous
N-2AO-V2H (V2H) voltage to current cards
have
failed due to design and/or component related
weaknesses.
CCW flow balance
tracked
under
50-315/316-96010-01.
NRC Inspection
Report
No. 50-315/316/95010,
Section 3.5, identified that the Unit
1 west
HDAFP had
a history of instantaneous
relay
(PJC
type) trips during Hode 4,
5 or 6 testing.
315/316-93012-Ola
5 c
IFI
Certain Westinghouse
protective action setpoints
did not include
an
EA term in diverse
secondary
(backup) trip functions.
315/316-93012-01d
315/316-93012-Ole
315/316-93012-01f
IFI
IFI
IFI
The
D.
C.
Cook design
used containment
pressure
to detect
a steam line break inside containment
rather than
steam line flow.
II
The Regulatory
Guide 1.97,
Category
1,
CST level
instrument
loop accuracy calculation did not
include
a seismic bias error term.
The
RWST level instrumentation
cables
were
routed through
an
HELB area.
10
IFI
IN
IPAP
MDAFP
SBICI
LIST OF ACRONYMS USED
Component
Cooling Water
Condensate
Storage
Tank
Current Transformer
Environmental
Allowance
Electrostatic
Discharge
Instrument
and Control
Inspection
Follow-Up Item
Information Notice
Integrated
Performance
Assessment
Process
Motor-Driven Auxiliary Feedwater
Pump
Motor Operated
Valve
Office of Nuclear Reactor Regulation
Reserve Auxiliary Transformer
Refueling Water Storage
Tank
Systems
Based
Instrument
and Control Inspection
Turbine Driven Auxiliary Feedwater
Pump
Updated Final Safety Analysis Report
Unresolved
Item
Violation