ML17321A484
| ML17321A484 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 02/27/1985 |
| From: | Wright G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17321A482 | List: |
| References | |
| 50-315-84-23, 50-316-84-25, NUDOCS 8503190133 | |
| Download: ML17321A484 (17) | |
See also: IR 05000315/1984023
Text
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U. S.
NUCLEAR REGULATORY CON>iISSION
REGION III
Reports
No. 50-315/84-23(DRP);
50-316/84-25(DRP)
Docket
Nos
~ 50-315;
50-316
Licenses
No. DPR-58;
Licensee:
American Electric Power Service Corporation
and >iichigan Electric Company
1 Riverside Plaza
Columbus,
OH
43216
Facility Name:
Donald C.
Cook Nuclear
Power Plant, Units
1 and
2
Inspection At:
Donald C.
Cook Site,
Bridgman, HI
Inspection Conducted:
December
22,
1984 through February
11,
1985
Inspectors:
B. L. Jorgensen
J.
K. Heller
Approved By:
G.
C. Wright, Chief
Reactor Projects. Section
2A
gl7 gS
Date
Ins ection
Summar
Ins ection on December
22
1984 throu h Februar
11
1985
Re orts
No. 50-315 84-23
DRP '0-316
84-25(DRP
Areas Ins ected:
Routine unannounced
inspection
by the resident
inspectors
of
licensee
actions
on previous inspection findings; operational safety; reactor
trips; surveillance;
Licensee Event Reports;
maintenance;
and independent
inspection activities.
The inspection involved a total of 242 inspector-hours
by two NRC inspectors
including 30 inspector-hours off-shift.
Results:
Of the seven
areas
inspected,
no items of noncompliance
or deviations
were identified in six areas;
one item of noncompliance
was identified in the
remaining area (failure to collect and analyze
compensatory
within required time limits - Paragraph
6).
8503
Cg 050003 1 5
190i33 850227
l
A
PDR)
A
DETAILS
1.
Persons
Contacted
-W.
E.
o'rB
- YP,
K.
- >A.
- J
>C.
G. Smith, Jr., Plant Manager
L. Townley, Assistant Plant Manager
A. Svensson,
Assistant
Plant Manager
A. Kriesel, Technical Supervisor,
Physical
Sciences
R. Baker, Operations
Superintendent
A. Blind, Technical Superintendent-Engineering
F. Stietzel,QC Supervisor
E. Murphy, Production Supervisor
The inspectors
also contacted
a number of licensee
and contract
employees
and informally interviewed operation, technical
and maintenance
personnel
during this period.
2.
>'<Denotes personnel
attending exit interview on February
19,
1985.
Licensee Action on Previous
Ins ection Findin s
(Open) Noncompliance
(315/81-28-01;
316/81-32-01
and 315/82-11-01;
316/82-11-01):
Noncompliance
315/81-28-01;
316/81-32-01
documented
that the controls for maintaining the "N" List as
an up-to-date,
reliable
reference
were inadequate.
Noncompliance
82-11-01 documented that while
verifying the adequacy of the updated
"N" List, the Divider Barrier Seal
(covered by Technical Specification 3/4.6.5.9
and considered to be safety
related)
was not listed.
This resulted in purchasing the seal
as
a non-
safety related item from a non-approved
vendor.
To correct these
items of
noncompliance the licensee
committed to review the "N" List for complete-
ness
and accuracy
and to make the required actions to make the list an
up-to-date
document.
The review and updating were scheduled to be com-
pleted by December
31,
1982.
While reviewing these
items the inspectors
were unable to find documentation that the review and updating had been
completed by December
31,
1982.
The licensee
was requested
to provide
documentation prior to the close of the next inspection report.
(Closed) Unresolved Item (315/82-15-07;
316/82-15-02):
This item ques-
tioned the Fire Protection Coordinator's practice of finding nonconforming
conditions
and using word-of-mouth to correct them.
Through discussion
with the Fire Protection Coordinator
and review of condition reports,
the
inspectors
found that nonconforming conditions are currently being identi-
fied on Condition Reports.
In addition, this item documented that findings
from the monthly
QC surveillance
housekeeping
reports
are not subject to
the condition reporting system.
Through discussion with the Quality Con-
trol Supervisor
and the Plant Manager,
the inspectors
were informed that
the housekeeping
Quality Control Surveillance reports
are reviewed by the
Plant Manager
who initiates corrective action.
(Closed)
Noncompliance
(315/82-22-04;
316/82-22-04):
The Plant Nuclear
safety Review Committee
(PSNRC) did not document that determination of an
unreviewed safety question
was
made for violations of Technical Specifica-
tions.
The inspectors verified the licensee action as stated in their
response
dated Narch 22,
1983.
No items of noncompliance or deviations
were identified.
0 erational
Safet,
Verification
The inspectors
observed control room operations,
including manning, shift
turnover,
approved procedures
and
LCO adherence,
reviewed applicable logs,
and conducted discussions
with control room operators
during the inspection
of December
22,
1984 through February
11,
1985.
Observations of control
room monitors, indicators,
and recorders
were
made to verify the operabil-
ity of emergency
systems,
radiation monitoring systems,
and nuclear
and
Reviews of surveillance,
equipment condition,
and tagout logs were conducted.
Proper return to service of selected
components
was verified.
Tours of the auxiliary building, turbine building,
Unit
1 containment,
and screenhouse
were made to observe accessible
equip-
ment conditions,
including fluid leaks, potential fire hazards,
and control
of activities in progress.
By observation
and direct interview the inspectors verified that the physi-
cal security plan was being implemented in accordance
with the station
security plan.
The inspectors
observed Unit 2 loose parts monitoring activities following
receipt of an alarm on the monitoring instrumentation during startup
on
January
3,
1985.
The licensee
and vendor
evaluation indi-
cated the likelihood that broken split pin parts,
as found in other plants,
were the cause of the condition.
The plant was cooled down, the steam
generators
opened,
and small split pin parts were found and retrieved
as
expected.
The licensee
subsequently
provided
NRC Region III a special
report on this matter, including a safety evaluation in support. of the
licensee's
review pursuant to
During a tour of the Unit 2 cable spreading
room, the inspectors
noted two
fire extinguishers
without current operability check tags attached.
Inde-
pendent
records verified the proper checks
had been performed
and current
tags were affixed to the subject extinguishers.
Unit 2 pressurizer
spray valve NRV-163, which contributed to a reactor
trip/safety injection on November
11,
1984 due to excess
leakage,
once
again became
suspected
of leakage despite intervening maintenance
and
tagging of the valve controller to maintain the valve in "manual" and
closed.
This was identified only after considerable
investigative effort
on the pressurizer
heaters
as
a potential source of pressure
control
problems - a situation contributed to by placement of the aforementioned
tag such that the valve controller was partially obscured.
The tag was
placed to the side of the controller and
a manual "closed" signal given
to the valve, which may have drifted slightly off its seat.
Several matters
came to the attention of the inspectors for which Licensee
Event Report's
(LERs) are anticipated.
These matters will be reviewed fully
in evaluation of the
LER information.
They include: for Unit 2 - identifi-
cation of a cold leg RTD installed with components
not all environmentally
qualified; for Unit 1 - concurrent inoperability during Mode 5 of both the
AB emergency diesel
and the
CD station battery without restoration of con-
tainment integrity within eight, hours; for both Units - identification of
multiple Rockwell RTD loop bypass isolation valve failures,
and; for both
Units - ice condenser
door testing to erroneous criteria such that surveil-
lance test failures were not identified and appropriate
LCO Action State-
ments
complied with.
Except for the Rockwell valve failure matter,
these
items
may have
NRC enforcement
action implications'his
was discussed
at
the Management
interview.
In review of preliminary information concerning concurrent inoperability
of the
AB diesel
and the
DC battery identified above,
the inspectors dis-
covered
a Technical Specification "interpretation" by the Plant Nuclear
Safety Review Committee
(PNSRC) involving LCO actions for inoperable
station batteries.
This interpretation
(No.
12 dated July ll, 1978)
appears
contradictory to Technical Specification 3.8 '.3, at least insofar
as no cross-reference
is made to Specification 3.0.3 for contemplated
cross-
tieing of the batteries
when one or the other battery is "inoperable".
This situation developed with Unit 2 in Mode
1 on January
30,
1985 when the
batteries
were cross-tied
under the provisions of 3.0.3 while a faulty cell
in one battery was jumpered.
An LER is forthcoming, on this matter,
which
will be reviewed further.
Concerning the
PSNRC interpretation,
the licen-
see
agreed to re-examine
the matter.
This was discussed
at, the Management
Interview.
Routine inspector review of Unit
1 logs identified an apparent
concurrent
inoperability of two engineered
safety feature fans, contrary to Technical
Specification requirements.
Further evaluation
and discussion with licen-
see personnel
established
that Specifications
had been met,
appearances
to
the contrary being
a consequence
of imprecise log entries.
The log entries
were corrected.
Observations
of the plant housekeeping/cleanliness
conditions
and the
implementation of the radiation protection program
and controls were made.
These reviews
and observations
were conducted to verify that facility oper-
ations were in conformance with the requirements
established
per Technical
Specifications,
10 CFR,
and Administrative Procedures.
No items of noncompliance or deviations
were identified.
Reactor Tri s
Following the plant trips discussed
below, the inspectors
ascertained
the
status of the reactor
and safety systems
by observation of control room
indicators and/or by discussions
with licensee
personnel.
The inspectors
verified the establishment
of proper communications
and reviewed the
corrective actions
taken by the licensee.
Unit 2 tripped from about 2/ power on January
12,
1985 when steam generator
lo-lo level developed
from a combination of steam
demand control difficulty
at low power and level shrinkage
due to increased
cold auxiliary feedwater
flow in response
to these difficulties.
System
responses
to the trip were
normal and the reactor
was
made critical at
1928 hours0.0223 days <br />0.536 hours <br />0.00319 weeks <br />7.33604e-4 months <br />.
Unit 2 tripped from about 96/ power on January
26,
1985 when power was lost
to a control room instrument panel
(GRID III) due to an internal computing
capacitor short circuit.
This caused
an indicated loss of reactor coolant
flow on one loop, tripping the reactor.
Subsequent
to the trip, the Tur-
bine Driven Auxiliary Feedwater
Pump
(TDAFP) failed to auto-start
on steam
generator lo-lo level.
Investigation revealed the
TDAFP would not stay
latched.
A governor linkage maladjustment
was corrected
and several
start tests,
including start
on the lo-lo level signal,
were performed
successfully.
Internal components of GRID III were replaced.
The Plant
Nuclear Safety Review Committee
(PNSRC) reviewed the circumstances
of the
trip and the findings concerning
GRID III and the TDAFP; restart
was
approved.
The reactor
was
made critical at 0241 hours0.00279 days <br />0.0669 hours <br />3.984788e-4 weeks <br />9.17005e-5 months <br /> the following day.
No items of noncompliance
or deviations
were identified.
Monthl
Surveillance Observation
The inspectors
reviewed Technical Specifications
required surveillance
testing on the systems listed below and verified that testing was performed
in accordance
with adequate
procedures,
that test instrumentation
was cali-
brated, that limiting conditions for operation were met, that removal and
restoration of the affected
components
were accomplished,
that test results
conformed with Technical Specifications
and procedure
requirements
and were
reviewed by personnel
other than the individual directing the test,
and
that deficiencies identified during the testing were properly reviewed
and
resolved by appropriate
management
personnel.
The following surveillance activities were observed/reviewed:
2
OHP 4030 STP.017
System Test
1
OHP 4030 STP.032
Quadrant
Power Tilt Ratio Calculation
12 THP 4030 STP.227
Multiple Entry Personnel Air Lock Leakage
Surveillance Test.
12 THP 4030 STP.204
Personnel Air Lock Leakage
and Interlock
Surveillance Test.
12 THP 6040 Per.091
1
OHP SP.032
RTD Bypass
Loop Flow Verification
Special Test of the Turbine Driven
Pump
No items on noncompliance or deviations
were identified.
6.
Licensee
Event
Re orts
Through direct observation,
discussions
with licensee
personnel,
and review
of records,
the following event reports
were reviewed to determine that
reportability requirements
were fulfilled, immediate corrective action was
accomplished,
and corrective action to prevent recurrence
had been
accom-
plished in accordance
with Technical Specifications.
The following LERs
are considered
closed:
Unit
1
RO 83-026/03L-0
RO 83-098/01T-0
and
1
RO 83-121/01T-0
Hole found in fire seal.
Incorrect data used for Cycle
7 Ep (z)
Backup sampling for vent stack monitor
not done.
RO 83-125/03L-0
RO 83-128/03L-0
T-Average exceeded
570
F.
AB Diesel Generator
due to
defective electronic tachometer circuit.
RO 83-130/03L-0
blowdown isolation
valve inoperable.
RO 84-001-0
Reactor trip January
23,
1984 during reactor
coolant loop flow instrument calibration.
An
incompletely closed equalizing valve on
Channel
3 (Instrument
NFP-222 which was in
"trip" for calibration)
reduced the pressure
differential across
associated
instruments
(common hi pressure
side tap) resulting in
apparent
low flow on a second
channel
( 2 of
3 to trip).
Plant response
was normal - see
IE Report 315/84-02.
Appropriate modifica-
tions were made to the subject calibration
procedures
(1 THP 6030.IMP.099 thru .101) to
ensure proper equalizing valve closure.
RO 84-002"0
On initial use of Rev.
11 of
1
OHP 4030
STP.005,
the Unit Supervisor discovered
an
error which could have rendered
both
trains inoperable,
in that the quarterly
portion of the test called for isolation of
one loop at a point in the procedure
where
the other loop was already isolated by the
monthly portion of the test.
The procedure
error was corrected.
RO 84-005-0
and
84-030-0
Required grab samples
were not taken within
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Table 3-13,
Item l.a, Action No.
30 when the auto gas analyzer
was inoperable.
In both cases
repairs
were started without
first obtaining
a grab sample,
and were in-
complete
(and preventing grab sampling)
when
the 12-hour time limit expired.
RO 84-026-0
and
84-029-0
Required grab samples
were not taken within
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per Table 3-13,
Item 3.a, Action No.
28 when the auxiliary building vent monitor
was inoperable.
In accordance
with NRC enforcement policy,
a licensee-identified
noncompli-
ance which should have been prevented
by corrective action for a previous
similar occurrence,
is subject to enforcement
action.
RO 84-029-0
and
RO 84-030-0 are each repeat
examples of previous similar occurrences
as
discussed
above;
and both represent
failure to comply with Technical Speci-
fication 3 '.3.10 '
in that required actions of Table 3.3-13
(ACTION 28 and
ACTION 30 respectively)
were not preformed within the times allowed.
Thus,
failure to collect and analyze required grab samples
in compensation
for
inoperable monitors
as described
above
and in RO 84-029-0
and
RO 84-030-0
is considered
an example of noncompliance with the referenced Technical
Specification for which a Notice of Violation is being issued with this
report.
(315/84-23-01
and 316/84-25-01).
RO 84-016-0
Motor driven auxiliary feedwater
(MDAFM)
pump handswitches
were in "neutral" during
Mode 3, making them incapable of responding
to selected
emergency auto-start signals.
A
noncompliance
(Item 325-84-18-02)
and esca-
lated enforcement
action (Civil Penalty)
relating in part to this matter,
have been
issued.
Corrective
and preventive actions
will be examined in followup of the noncom-
pliance.
RO 84-017-0
MDAFW pumps started automatically
as designed
when the single operating main feed
pump was
lost during startup.
Manual steam generator
level control involving both an operator
and
a trainee resulted inadvertently in high S/G
No.
4 level.
RO 84-019-0
The turbine driven auxiliary feedwater
pump
governor valve position was left in a condi-
tion per instructions of a surveillance
such
that it was not capable of meeting design
pressure
and flow requirements.
This item
was part of a noncompliance
(Item 315/84-18-
03 and 316/84-20-01)
and escalated
enforce-
ment (Civil Penalty).
Corrective
and preven-
tive actions will be examined in followup of
the noncompliance citation.
RO 84-024"0
Spent fuel exhaust ventilation dampers did
not initially respond to input of a high
alarm on radiation monitor R-5 during a test.
No system problem could be found; the dampers
responded
properly on subsequent
tests
(the
event could not be repeated)
and were left
"operable" after testing.
RO 84-025-0
System walkdowns being conducted
under the
D.
C.
Cook Regulatory Performance
Improvement
Program identified mislabeled firewater ring
header valves
as such required testing was
being applied to the wrong valves.
Proce-
dures
were corrected
and the correct valves
were tested satisfactorily.
Unit 2
RO 83-042/03L-0
RO 83-069/03L-0
l
Hydrogen sampling system inoperable
Component cooling water pump discharge valve
RO 83-102/03L-0
Non essential
service water valve from lower
containment, ventilation Unit No.
3 closing
time.
RO 83-117/03L-0
Number
blowdown sample
valve inoperable
due to a damaged actuator.
RO 84-001/0
A boron injection tank (BIT) was diluted
when, following maintenance,
the system line-
up restoration did not address
the valve
which had been worked on and it was left
partially open.
The
LCO Action Statement
(1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restoration)
was met.
RO 84-012-0
Inadvertent
Node
6 Safety injection signal
(no injection) due to insufficient specific
test procedure instructions for sequence
of
returning components
(solid state protection
system) to service.
The procedure
was
corrected.
RO 84-013-0
This was
a voluntary licensee report concern-
ing the
AB battery 60-month capacity
(discharge) test,
which showed the battery
had reached
the end of its useable life.
The
battery was replaced
and the new battery
tested
and placed in service.
RO 84-014-0
Both RHR pumps were inoperable in diode
5 for
about
25 minutes
when operators
switching
from one
pump to the other started the second
pump before stopping the first while in "half-
loop" operation.
Both pumps
became
airbound
and had to be vented before being returned
to service.
The procedure
was revised to
clarify that restrictions
on two-pump opera-
tions apply even during the brief switching
process.
RO'84-015-0
RO 84-017-0
'urveillance
requirements
of Technical Speci-
fication 4.0.2.6 were not met when
a required
was analyzed
35 minutes late.
The technicians
involved were not adequately
informed concerning either the sample loca-
tion or the deadline.
They were reinstructed
and the sample
frequency
was increased to
prevent recurrence.
Shiftly channel
checks of the intermediate
range excore detectors
were not performed
in accordance
with Technical Specification
Table 4.3-1 (asterisked notation) for a
condition other than Nodes
1 or
2 with the
reactor trip breakers
closed
and the rod
drive system
capable of rod withdrawal.
The
surveillance procedure
(2
OHP 4030 STP.030)
was revised to include the asterisk notation
requirements.
RO 84"018"0
Control rods B-8 and K-10 indicated
14 and
13
step misalignments,
respectively,
from their
group demand positions
- the limit being
12
steps.
The applicable Action valuation indi-
cated the secondary coil stack measurements
used for position determination
had not yet
stabilized following thermal cycling associ-
ated with the post-refueling
low power physics
testing in progress
at the time.
Thus, the
"misalignments" were likely an indication
error.
The rods indicated proper alignment
on withdrawal the following day.
RO 84-019-0
and
84-019-1
When a reactor coolant loop was
removed from
service
and T-avg maintained
above
541
F,
associated
loop bistables
were not tripped
within one hour per Technical Specification 3.3.2-1, Table 3.3-3,
Item 4d.
This matter
was inspector-identified
and
a noncompliance
citation (Item 316/84-14-02)
issued.
Correc-
tive action will be reviewed in followup to
the noncompliance citation.
4t
RO
RO
84-024-0
84-025-0
84-031-0
84-032-0
These reports all involve Unit 2 low power
or zero power
RPS actuations
(trips) which
have been previously reviewed
- the first
two in IE Report 316/84-21,
the other two
in IE Report 316/84-24.
One item of noncompliance
was identified involving a repetitive failure
to collect timely compensatory
samples
when automatic
equipment
was
7.
Inde endent
Ins ection Activities
Licensee Condition Reports
were routinely reviewed to provide an on-
going and current perception of the effectiveness
of the licensee's
corrective action program.
Selected
reports
were reviewed in more
detail on the basis of their potential implications or the particular
interest of the inspectors.
Condition Report, 2-12-84-2641
discussed
lifting of an
RHR safety valve during Unit 2 airsweeps
on December
23,
1984.
Procedure
2
OHP 4021.002.001
was used in a manner
(higher
'ressure)
not contemplated
and,
as such, this event could be consid-
ered
a consequence
of violating an approved safety-related
procedure.
Condition Report 1-12-84-2598
involved "E" motor driven auxiliary feed-
water
pump emergency leakoff valve inoperability as
a consequence
of
violating procedural controls
on lifted leads.
Under
NRC Enforcement
Policy, minor violations of this type identified and corrected
by the
licensee
are not subject to NRC Enforcement Action.
b.
The licensee's
response
to IE Bulletin 8'g-03 "Refueling Cavity Mater
Seal"
was subjected to preliminary review.
Copies of this response
distributed to the site were apparently transmitted without an oath
or affirmation statement,
though the copy sent to
NRC Region III had
such
a statement.
The inspectors
requested
a copy of the affirmation
statement
be transmitted to the site.
Item II.B.l "Reactor Coolant System Vents" and
10 CFR 50.44 (c)(3)(iii) provided for installation and operability (including
procedure preparation
and operator training) of RCS vent systems for
use in controlling potential noncondensable
gas accumulation in the
The licensee installed the vent system
as
discussed
in IE Report 315/84-20
and 316/84-22,
leaving "open" the
completion of surveillance testing,
procedure revisions, training,
and
conversion of the systems to full operability, pursuant to newly
issued Technical Specifications.
During unit outages for each of the respective units during this
inspection period, the licensee verified or performed current sur-
veillance testing
and energized
and declared
the respective
vent
systems
"operable".
Procedure
revisions covering system operation
were approved
and issued
(TP-1 to Rev.
1 dated
May 15,
1984 of Proce-
dures
1 and
2
OHP 4023.001.015) prior to the operability declarations.
Operator training was performed after installation of the actual modi-
fications with a refresher
memorandum to the shifts
(October
19,
1984)
relating to final elevation of the systems
to full operability under
the Technical Specifications.
10
The inspectors,
having no further questions
concerning this item, con-
sider this item closed.
d.
The inspectors
discussed
internal
NRC information potentially appli-
cable to D. C.
Cook with appropriate
licensee representatives.
On
January
12,
1985,
a reactor trip breaker at Sequoyah Unit 2 did not
open automatically
on demand
due to a printed circuit board short
(i.e., the breaker did not "fail safe").
The licensee
was provided
with available information and asked to review the matter for appli-
cability.
On a separate
matter, the licensee
was requested
to review
his records for possible historical procurement of safety-related
com-
ponents
from a vendor (Familian Northwest) under investigation for
falsification of material certifications.
The licensee
review indi-
cated this vendor has not supplied safety-related
components to
D.
C. Cook.
No items of noncompliance or deviations
were identified.
8.
Mana ement Interview
A management
interview (attended
as indicated in Paragraph
1) was conducted
at the completion of the inspection.
The following items were discussed:
a.
The inspectors
summarized the scope
and findings of the inspection
as
described in these details.
b.
The apparent
item of noncompliance (partially affecting both Units)
was specifically identified and discussed
(Paragraph
6).
c.
Events for which LERs (or supplemental
information thereto)
are anti-
cipated
and which may have
NRC enforcement
action implications were
identified and discussed
(Paragraph
3).
d.
A questionable
PNSRC Technical Specification "interpretation" concern-
ing cross-tieing station batteries
was discussed
- the licensee
agreed
to re-examine
the matter
(Paragraph
3).
e.
The TMI Action Item (NUREG-0737 Item II B.l) involving reactor coolant
vents
was identified as "closed" on the basis of this inspection
(Paragraph 8.c).
The inspector
asked the licensee
representatives
whether they
considered
any of the matters
discussed
to contain proprietary
information or other information exempt
from disclosure pursuant to
No such information was identified.
11
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gr