ML17324A711
| ML17324A711 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/11/1986 |
| From: | Mccormickbarge, Ring M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17324A710 | List: |
| References | |
| 50-315-86-06, 50-315-86-6, 50-316-86-06, 50-316-86-6, NUDOCS 8603170368 | |
| Download: ML17324A711 (11) | |
See also: IR 05000315/1986006
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION III
Reports
No. 50-315/86006(DRS);
50-316/86006(DRS),
Docket Nos.
50-315;
50-316
Licenses
No.
Licensee:
American Electric Power Service Corporation
and Michigan Power
Company
1 Riverside Plaza
Columbus,
OH
43216
Faci l ity Name:
D.
C.
Cook Nucl ear Pl ant,
Units 1 and
2
Inspection At:
D.
C.
Cook Site,
Bridgman,
MI
Inspection
Conducted:
February 3,
1986 through February 26,
1986
Inspector:
M.
L. McCormick-Barger
5-6-sk
Date
Approved By:
M. A. Ring, Chief
Test Programs
Section
'5-<l<<t4
Date
Ins ection
Summar
Ins ection
on Februar
3
1986 throu
h Februar
26
1986
Re orts
No.
50-315/86006(DRS
. 50-316 86006(DRS
previous inspection findings, licensee
event reports
estimated critical
condition calculation for a Unit 2 startup
on February 3, 1986, core thermal
power, Unit 1 Cycle 9 rod worth measurements,
Unit 1 Cycle
9 moderator
temperature coefficient measurement,
and core power distribution limits.
The
inspection involved 54 inspector-hours
onsite
and two inspector-hours offsite by
one
NRC inspector.
Results:
No violations or deviations
were identified.
Bb03i703b8
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8
PDR,
,DETAILS
1.
Persons
Contacted
A. Blind, Assistant, Plant Manager
R. Baker, Operations
Superintendent
S.
Gibson, Technical-Engineering
Department
Head
W. Hennen,
Nuclear Engineering Supervisor
Verteramo,
Nuclear Performance
Engineer
F. Stietzel, guality Control Superintendent
J.
Nadeau,
AEPSC guality Assurance
"Denotes
those attending the exit interview on February 26, 1986.
2.
Licensee Action on Previous
Ins ection Findin
s
a 0
b.
C.
(Closed)
Unresolved
Item (316/84-16-01):
Appendix A-205 of
Procedure
2 THP 4030.
STP.330,
Revision 5, "Surveillance of Core
Distribution Limits," contained incorrect Technical Specification
acceptance
criteria.
During this inspection,
the inspector
reviewed
Procedure
2 THP 4030.
STP.330,
Revision 7,
and the corresponding
procedure for Unit 1 (1THP 4030.
STP.330,
Revision 9) and found that
the Technical Specification limits, contained in Appendix A, for each
of these
procedures
agreed with the Unit 2 and Unit 1 Technical
Specifications,
respectively.
The inspector
had
no further concerns.
(Closed)
Unresolved
Item (316/84-16-02):
The procedure results for
Procedure
12
THP 6040. PER. 350, Revision 0, "Isothermal
Temperature
Coefficient Measurement
and Moderator Temperature Coefficient
Calculation," performed for the Unit 2 Cycle
5 startup
lacked
documentation
related to some unanticipated
slope
changes
in the
reactivity versus
temperature plots.
During this inspection,
the
inspector
reviewed the licensee's
actions which included
documentation
of the licensee's
evaluation pertaining to the
unanticipated
slope
changes
and
a revision to Procedure
12 THP
6040. PER. 350.
A note following Step
8. 6. 1 of Revision
1 of this
procedure
discusses
deviations
from a straight line for an
isothermal
temperature coefficient measurement
plot.
Based
on the
way this note was written, the inspector
was concerned it might be
misinterpreted to mean that,
should unusual plots occur in the
future,
no pursuit of the cause
would be necessary.
The licensee
made
a commitment to clarify the procedure prior to the Unit 2
Cycle
6 startup to ensure that an evaluation would be performed if
slope irregularities occurred in future tests.
This is considered
an open item (315/86006-Ol(DRS);
316/86006-01(DRS))
pending
procedure revision and subsequent
NRC review.
(Closed) Unresolved
Item (316/84-16-03):
During the performance of
Procedure
12 THP 4030.STP.307,
"Moderator Temperature Coefficient
Determination," for the Unit 2 Cycle
5 startup,
Step 8.2 and the
associated
data sheets
were
mar ked "N/A" not applicable with no
justification given.
The use of "N/A" was not permitted without a
change
sheet at the time this step
was marked "N/A."
However,
based
on the inspector's
review of the corrective action which included
documentation of the justification for the "N/A," the justification
was reasonable
and'here
was
no safety significance to skipping this
step
and the associated
data sheets.
In addition, generic
corrective actions
were taken.
Several
Plant Manager Instructions
(PMIs) were revised to address
the use of "N/A" during performance
of procedures
including PMI-2010, "Plant Manager
and Department
Head
Instructions,
Procedures,
and Associated
Indexes,"
and PMI-6040,
"Performance/Engineering
Test Procedures."
Step 3.1. 2 of PMI-2010
stated,
in part, that "It is expected that as procedures
are
revised,
instructions for partial completion will be addressed
on a
case
by case basis."
The inspector discussed
this with the Nuclear
Engineering Supervisor
who indicated that the procedures
used
by the
Nuclear Group had been
reviewed and,
where appropriate,
revised to
identify procedure
steps that are optional
under certain conditions
and can, therefore,
be marked "N/A."
The inspector
reviewed
numerous
examples of procedures
used by the Nuclear Group which had
been revised to address
the use of "N/A."
There
was
a statement
that appeared
in several
Nuclear Group procedures
(for example:
"~THP 6040.PER.356,
"Reactivity Computer Checkout" ) that concerned
the inspector
because it might be misinterpreted
as giving unlimited
authority to Test Engineers
as
opposed to controlling the use of
"N/A" through procedural
allowances
as described
in PMI 2010.
The
statement
was:
"Only those sections
deemed appropriate
by the test
engineer
need
be completed for each specific test."
The licensee
committed to look into clarification of that statement.
This is
considered
an open item (315/86006-02(DRS);
316/86006-02(DRS))
pending licensee action and subsequent
NRC review.
No violations or deviations
were identified, however,
two areas
require further review and will be followed as
open items.
3.
Licensee
Event
Re orts
Through discussions
with licensee
personnel
and review of records,
the
following Licensee
Event Report
(LER) was reviewed to determine that
reportability requirements
were met,
and corrective
and preventive
actions
were accomplished
in accordance
with Technical Specifications.
The following LER is considered
closed:
LER 315/84009:
A flux map taken at 99K power on March 21,
1984,
indicated that the heat flux hot channel factor (F~) approximately
0.4X.
Power level
was subsequently
reduced to 96K power.
Reanalysis
demonstrated
that this was
an indicated rather
than
actual violation based
on conservatisms
incorporated in the original
flux map analysis.
The reanalysis
indicated that it would be
permissible to return to 99.7X power.
Further reanalysis utilizing
a burnup dependent
V(z) indicated that it would be possible to
return to 103X power without violating the
F~ Technical
1
Specification limit.
Power was then returned to lOOX power.
Since the Technical Specification limit was not actually violated,
this
LER was submitted
as
a Voluntary LER.
The inspector
reviewed the following records:
Unit 1 Cycle 8 Flux Map 108-35,
March 21, 1984, Option No. 6, Job
No.
1084,
2-D Analytical Factors (with 3X penalty applied to
F ).
Unit 1 Cycle 8 Flux Map 108-35,
March 21, 1984, Option No. 6, Job
No. 979,
3-D Analytical Factors,
V(z) was not core burnup dependent.
Unit 1 Cycle 8 Flux Map 108-35,
March 21, 1984, Option No. 6, Job
No. 1601,
2-D Analytical Factors
(with 1X penalty applied to
F ),
core burnup dependent
V(z) factor, Control
Bank
D at 216 steps.
Unit 1 Cycle 8 Flux Map 108-36,
March 27, .1984, Option No. 3, Job
No. 2546,
3-D Analytical Factors,
Control
Bank
D at 220 steps.
Unit 1 Control
Room Log Book 24, March 21 through 25, 1984.
The inspector
had
no concerns
based
on the review.
No violations or deviations
were identified.
4.
Estimated Critical Condition Calculation for a Unit 2 Startu
on
Februar
3
1986
During a Unit 2 reactor startup
on February 3, 1986, the reactor failed
to achieve criticality within the limits predicted
by the estimated
critical condition calculation.
The inspector discussed this with licensee
personnel
and reviewed the following records:
Condition Report 2-02-86-132, "Failure to Achieve Criticality on
Unit 2 Reactor
as Predicted
by the Estimated Critical Condition
Calculation," initiated on February 3, 1986.
Procedure
2-OHP 4021.001.011,
Revision 2, (with Change
Sheet
No.
1
attached),
!'Determination of Critical Conditions."
Procedure
2-OHP 4021.001.002,
Revision 7, "Reactor Start-up."
Data/Signoff Sheets for Procedure
2-OHP 402.001.011,
"Determination of
Critical Conditions," performed
on August 16,
1985, October 31, 1985,
February 1, 1986 and February 3,
1986 (recalculation following the
Unit 2, February 3, 1986, startup termination).
Individual Training Records for the nuclear performance
engineer
who
preformed the estimated critical conditions calculation
on February 1,
1986.
Unit 2 Control
Room Log Book for February 3, 1986.
t
M
V
A 1/m plot is performed for every Unit 2 startup in accordance
with
Procedure
2
OHP 4021.001.002,
"Reactor Startup,"
and,
using this plot, the
reactor operators
determined that criticality would not occur within the
limits of the Estimated Critical Condition (ECC) calculation.
The reactor
startup
was terminated at 6: 51 a.m.,
and all control
banks were fully
inserted
by 7:05 a.m..
The licensee's
investigation
showed that a value was
used in the
calculation which was obtained,off of an inappropriate
curve from a Unit 2
Technical
Data Book figure.
(The figure contained
several
curves
and,
whereas
the
ECC calculation procedure
specified the figure number, it did
not specify which of the curves
should
be used).
Although the curve used
was not the one typically used, it was conservative
and, therefore,
was
accepted
by the two Senior Reactor Operators that signed off on the
calculation.
The inspector
reviewed the training records of the
individual that performed the
ECC calculation.
These
records
showed that
the individual had received training for the "Determination of Critical
Conditions" procedure.
In addition, the individual had performed
an
calculation
on two previous occasions
(August 16,
1985 and October 31,
1985)
and on both of these
occasions
the calculation
was performed
correctly and had been
checked
by another
member of the Nuclear Group.
After determining the cause of the error in the
ECC calculation,
the
licensee
recalculated
the estimated critical conditions
and, during the
reactor startup that followed, went critical within the limits of the
calculation.
The licensee's
actions to prevent future errors related to
the selection of an improper curve from a Technical
Data Book figure
included, revisions of Units 1's
and Unit 2's
ECC calculation procedures
to identify the specific curve that should
be used.
In addition, the
licensee
reviewed other operations
and nuclear performance
procedures
to
determine if similar revisions
were warranted in other procedures.
As a
result of this review, the licensee
intended to revise
one additional
procedure
(12 THP 4030.STP.308,
"Boron Curve Update" ).
Based
on the
licensee's
corrective actions
and the fact that this appeared
to be an
isolated occurrence,
the inspector
has
no further concerns.
No violations or deviations
were identified.
Core Thermal
Power
The inspector
reviewed licensee
procedures
and results to verify that the
calculation of core thermal
power was technically correct
and that
results indicated that reactor power was within prescribed limits.
The
inspector utilized the following documents
during the review:
Procedure
1 OHP 4030. STP. 029, Revision 8, "Reactor
Thermal
Power
Determination."
Procedure
2
OHP 4030.STP.029,
Revision 5, "Reactor Thermal
Power
Determination."
I
I4
lf
1[
e
Procedure
12 THP 4030.STP.219,
Revision 3, "Thermal
Power
Measurement
and Reactor
Coolant System
Flow Rate," performed for
Unit 1 Cycle 9, October 23,
1985 through January
18,
1986.
Data/Signoff Sheet
6. 1 for Procedure
1 OHP 4030.STP.029,
Revision 7,
"Reactor Thermal'ower Determination'
for the following dates:"
January l-ll, 1985
a'nd January
29 through, April 5,
1985.'ata/Signoff
Sheet 6.3 for Procedure
1 OHP 4030.STP.030,
Revision 11,
"Operations Shift Surveillance
Checks, (Modes 1, 2, 3, 4), for the
following dates:
February
15 through March 31, 1985.
Computer Printout - Daily Trend, Block Data for Unit 1, March 1985.
No violations or deviations
were identified.
6.
Unit 1
C cle
9 Rod Worth Measurements
The inspector
reviewed the Unit 1 Cycle
9 rod worth measurement
procedure
for technical
adequacy
and verified that the, results satisfied the
acceptance
criteria.
The following documents
were used in this review:
Procedure
1 THP SP. 101, Revision 0,
"Rod Worth Verification Tests
Utilizing RCC Bank Interchange,"
dated October 8,
1985 and performed
for Unit 1 Cycle
9 on November 15,
1985.
"Core Physics Characteristics
of the Donald
C.
Cook
Station Nuclear Plant (Unit 1, Cycle 9)," August 1985.
No violations or deviations
were identified.
7.
Unit 1
C cle 9 Isothermal/Moderator
Tem erature Coefficient
The inspector
reviewed the Unit 1 Cycle
9 low power physics test
procedure
related to isothermal
and moderator temperature coefficient
measurement
for technical
adequacy
and verified that the results
satisfied the acceptance
criteria.
The following documents
were used in
this review:
12
THP 6040. PER.350,
Revision 1, "Isothermal
Temperature Coefficient
(ITC) Measurement
and Moderator Temperature Coefficient (MTC)
Calculation," dated April 15,
1985 and performed for Unit 1 Cycle 9
on November 14,
1985.
"Core Physics Characteristics
of the Donald
C.
Cook
Station Nuclear Plant (Unit 1, Cycle 9)," August 1985.
Comments related to this review are contained in Par'agraph
2.b.
No violations or deviations
were identified.
1
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I
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4
'
Core Power Distribution Limits
The review of licensee
procedures
and results related to core power
distribution limits for Unit 1 Cycle 8 and the startup of Cycle
9 began
during
NRC Inspection 315/85034;
316/85034 (refer to Paragraph
5 of
Inspection
Report
No. 315/85034;
316/85034 for details).
During this
inspection the inspector
reviewed
one additional
document related to core
power distribution limits:
Technical Specification Clarification No. 17, Technical
Specification
3. 2. 4, "Application of the quadrant
Power Tilt
Technical Specification,"
reviewed by the Plant Nuclear Safety
Review Committee
on January 8, 1980.
The Technical Specification Clarification identified above, will be
referred to NRC's Office of Nuclear Reactor Regulation
(NRR) for further
review and evaluation.
This will be followed as
an open item
(315/86006-03(DRS);
316/86006-03(DRS))
pending completion o'f NRR's review.
No violations or deviations
were identified, however,
one item requires
further review and will be followed an an open item.
9.
~0en Items
Open items are matters
which have
been discussed
with the licensee,
which
will be reviewed further by the inspector,
and which involve some action
on the part of the
NRC or licensee
or both.
Open items disclosed
during
the inspection are discussed
in Paragraphs
2.b, 2.c and 8.
10.
Exit Interview
The inspector
met with licensee
representatives
(denoted in Paragraph
1) on
February 26,
1986 to discuss
the scope
and findings of the inspection.
The
licensee
acknowledged
the statements
made
by the inspector with respect to
items discussed
in the report.
The inspector also discussed
the likely
informational content of the inspection report with regard to documents
or
processes
reviewed
by the inspector during the inspection.
The licensee
stated that portions of the rod bank interchange
procedure
(1 THP SP. 101)
and the core physics characteristics
report
(MCAP-10862) were considered
proprietary but that references
to them would not be.