ML17321A770
| ML17321A770 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 08/08/1985 |
| From: | Eng P, Milbrot W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17321A769 | List: |
| References | |
| 50-315-85-21, 50-316-85-21, IEB-84-03, IEB-84-3, NUDOCS 8508130253 | |
| Download: ML17321A770 (16) | |
See also: IR 05000315/1985021
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION III
Repor t No. 50-315/85021(DRS);
50-316/85021(DRS)
Docket No. 50-315;
50-316
License
No. DPR-58;
and
Licensee:
American Electric Power Service Corporation
and Michigan Power
Company
1 Riverside
Plaza
Columbus,
Ohio 43216
Facility Name:
D.
C.
Cook Nuclear Plant, Units
1 and
2
Inspection at:
D. C.
Cook Site,
Bridgman, Michigan
Inspection Conducted:
July
15 through
18,
1985
Inspectors:
P. L. Eng
Date
M. E. Miibro~ g.
a
Approved By:
ll. G. Guldemond,
ief
Operational
Programs
Section
Da
e
ate
Ins ection
Summar
Ins ection
on Jul
15 throu
h Jul
18
1985
Re ort No. 50-315/85021
DRS;
RS
reas
ns ected:
outine,
announced
inspection of licensee
actions
on previous
inspection fin ings; inservice testing
program for valves; refueling activities;
and licensee
actions
regarding
The inspection
involved a
total of 50 inspector -hours onsite
by two
NRC inspectors.
Results:
Of the four areas
inspected,
one violation was identified (failure
Co foCTow procedures
- Paragraph
4).
e
8508i30253 850808
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DETAILS
1.
Persons
Contacted
- W.
- J
- K.
- N
T.
- A.
- J
- C
- L
M.
T.
- M
- T
- R
- T
- J
- M.
- S
G. Smith, Jr., Plant Manager
D. Allard, Maintenance
Superintendent
R. Baker, Operations
Superintendent
Baker, guality Control Department Assistant
P. Beilman, Planning Supervisor
A. Blind, Assistant Plant Manager
R. Bobay, Project Superintendent
- Planning
A. Freer, guality Control/ Inservice Inspection
S. Gibson, Technical
Superintendent
L. Horvath, guality Assurance
Supervisor
Kriesel, Technical
Superintendent
- Physical
Sciences
A. Lester, Senior Performance
Engineer
K. Postlewait,
Performance
Supervisor
Simms, Station Superintendent
R. Stephens,
Performance
Engineer - Operations
F. Stietzel, guality Control Superintendent
S. Ackerman, Nuclear Safety
5 Licensing,
AEPSC
A.
Mc Aligott, guality Assurance Auditor, AEPSC
- Denotes those attending
the exit interview held
on July 18,
1985.
Additional plant technical
and administrative
personnel
were contacted
during the course of the inspection.
2.
Action of Previous
Ins ection Findin
s
a
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b.
C.
(Cl osed) Violation (315/84-13-Ol(DRS);
316/84-15-01(DRS) )
Turbine
Driven Auxiliary Feed
Pump
(TDAFP) discharge
pressure
allowed to
violate Technical Specification
(TS) limits.
The licensee
has
submitted
a proposed
TS change clarifying temperature
bases
for
TS
limits on the
TDAFP and deleted
the provision for temperature
compensation
from the
TDAFP test procedure.
(Open)
Unresolved
Item (315/84-13-03(DRS);
316/84-15-03(DRS) ):
Response
time testing of the turbine driven auxiliary feedwater
pump
per technical specification requirements.
This item remains
open
as
the licensee
has not identified the test method to be used.
The
licensee
indicated that testing would be performed in conjunction
with the
TDAFP test scheduled
during the startup of Unit 1 and the
next scheduled
quarterly
TDAFP test for Unit 2.
(Closed)
Unresolved
Item (315/84-12-03(DRS);
316/84-14-03(DRS));
96
hour operability determination.
The licensee
has incorporated
the
action limits for the inservice testing of pumps into the test
procedures
and inserted operability limits into the technical
data
book located in each unit's control
room, thereby providing the data
immediately following completion of
surveillance testing.
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d.
(Closed) Violation (315/84-13-06(DRS);
316/84-15-06(DRS)):
Failure
to implement
an inservice testing
program conducive to identifying
conditions adverse to quality.
The licensee
has taken the following
actions in response
to the violation:
(1)
Limiting stroke times for valves are being revised.
The
licensee
has identified alert and action times in accordance
with Code requirements for valves
based
on the first four
inservice test times
obtained following initial implementation
of the inservice testing program.
The licensee
stated that
these
times will be used for component operability determination
by December
31,
1985.
(2)
The licensee
has revised its valve stroke time records to
include all stroke time data including post modification and
maintenance
data.
In addition,
an increased
frequency log has
been established
to document those valves which are tested
on
increased
frequency.
A matrix of like valves to be used for
evaluation of generic
concerns
is being prepared
and will be in
use
by December
31,
1985.
Program level documentation for valve
problems is addressed
in licensee
procedure
12-gHP-5070ISI.014,
"ISI Valve Data Recording
and Corrective Action for Power
Operated
Valves," Revision 0, dated April 1, 1985.
The inspector
provided
an information copy of an Office of Nuclear Reactor
Regulation
(NRR) memo, attached,
clarifying the
NRC interpretation
of maximum stroke time requirements for inservice testing of
valves.
(3)
The licensee
stated that
a fixed set of maximum valve stroke
time alert and action ranges,
which are in accordance
with Code
acceptance criteria, will be used to determine
the status of the
valve in lieu of the percent
increases
of valve stroke times
as
defined in Section
XI of the American Society of Mechanical
Engineers
(ASME) Boiler and Pressure
Vessel
Code,
1974 Edition
including the appropriate
addenda.
Changes
to these
acceptance
criteria will be evaluated
in a manner similar to that described
in Section
XI for pump reference
value changes.
Licensee
procedure
12-(HP-5070ISI.014
requires that infrequently tested
valves exhibiting unacceptable
increases
in stroke times
be
evaluated
or repaired prior to mode change of the affected unit.
This is acceptable.
With regard to leak test trending, the licensee
stated that they
have requested relief from the trending requirements for
containment isolation valves (CIV); however, trending will be
performed for leak tested
valves which are not CIVs.
This
matter is discussed
further in Paragraph
3.
The licensee
has effectively addressed
the concerns identified by
the violation.
e.
(Open) Unresolved
Item (315/84-13-07(DRS);
316/84-15-07(DRS) ):
Remote position indication verification of all valves.
The licensee
responded
to this item by letter dated April 12,
1985, indicating
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that remote position verification for Unit 1 valves will be completed
prior to the end of the current outage; Unit 2 valves will be
completed
by the end of the unit's next scheduled
outage.
The
inspector
expressed
concern that procedures for verifying remote
position indicators for accessible
valves
have not been written.
The licensee
maintains that remote position verification for Unit
1
valves will be completed prior to the end of the current outage;
Unit 2 valves will be completed
by the end of the unit's next
scheduled
outage.
This item remains
open pending completion of the
licensee's
commitment in their April letter.
(Open)
Open
Item (315/84-13-08(DRS);
316/84-15-08(DRS) ): Recording
first stroke for valve timing.
The licensee
has
issued
a
memo to
equipment operators
and operations staff requiring that the first
stroke time for a valve be used for operability determination.
In
addition, the licensee
is in the process
of incorporating the
requirement into all valve stroke time procedures.
This item
remains
open pending inspector review of the procedures
and test
data.
(Open)
Open
Item (315/84-13-09(DRS);
316/84-15-09(DRS)):
Valve
stroke time limits not provided to maintenance for post modification
testing.
The licensee
stated that the method of providing the
revised stroke times discussed
in Paragraph
2.d.(1) for post
modification/ maintenance
testing
has not yet been determined
but
will be accomplished
by December
31,
1985. Availability and review
of post maintenance/
modification test data will be reviewed in
subsequent
inspections.
(Closed)
Open Item (315/84-13-10(DRS);
316/84-15-10(DRS) ): Review of
valve leak test data for validity.
The licensee
has revised
procedure
12 THP 4030.STP.226,
"Surveillance Test Procedure
Check Valves," to include
a precaution
statement
addressing verifica-
tion of zero valve leak rates with regard to magnetically coupled
rotameters.
The licensee
has evaluated
previous valve leak test data
and found it acceptable.
(Closed)
Open
Item (315/84-13-12(DRS);
316/84-15-12(DRS)):
Correlation of January
bearing
temperatures
to summer conditions.
The licensee
has
rescheduled
inservice testing annual
bearing
temperature
measurements
to August.
(Closed)
Open item (315/84-13-13(DRS);
316/84-15-13(DRS)):
Use of
acceptance
ranges for both flow and pressure
during inservice
testing of pumps.
The licensee
stated that due to flow anomalies,
achieving the exact reference
value flow was extremely difficult,
and flow and the corresponding
observed
pressure
used to determine
pump operability were subject to the limitations and instrument
requirements
delineated
in the
ASME Code.
The inspector
reviewed the
licensee's
records
and correction factor
used for correcting test
pressures
based
on observed
flow and found them acceptable.
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Inservice Testin
Pro
ram for Valves
During the review and discussions
with members of the licensee's
staff
regarding closure of the violation discussed
in Paragraph
2.d of this
report, the inspector noted that the licensee
had not addressed
trending
and evaluation
requirements for valve leak testing
as delineated
in Sub-
sections
IWV-3420f and
IWV-3420g of Section XI.
The inspector provided
a
copy of a
memo from NRR, attached,
regarding valve leak testing
and stated
that the requirements
of IWV-3420f and
IWV-3420g apply.
The licensee
repre-
sentatives
stated that they would evaluate
the position stated in the subject
memo and seek further clarification from NRR on this subject.
Implementation
of valve leak test trending by the licensee
per the
Code requirements
or the
granting of relief from said requirements
by the Commission will be tracked
as
an open item (315/85021-01(DRS);
316/85021-01(DRS)).
No violations or deviations
were identified.
Refuel in
Pre arations
The inspector
reviewed procedures,
tests
and surveillances
covering the
maintenance,
testing
and operational
check out of refueling tools,
equipment
and systems
required to support the fuel loading effort to
assure
that the applicable Technical Specifications
have
been included.
Equipment
and components
to be used during the performance of refueling
activities were checked for proper operation
and verified ready for use.
The inspector also reviewed completed surveillances
that had to be met
prior to entry into Mode 6.
The surveillances
were completed
as required.
Fuel handling personnel
training was completed
as required
and results
documented
on
a qualification letter and personal
work experience
records.
The inspector
reviewed several
completed refueling procedures for core
alteration preparations.
The review included recording of required data
and verification sign offs.
Two of the procedures
reviewed were
Refueling Procedure
FP-AEP-R8, Revision 8, paragraph
9.2.6,
"Reactor Cavity Seal
Ring Installation
and Removal,"
and paragraph
9.2.8,
"Reactor Vessel
Head
Removal
and Installation."
Two record copies
are
maintained of the Westinghouse
refueling procedures.
When the refueling
work area is within a contaminated
area,
a working copy of the procedure
is provided at the job site
and the two record copies
are maintained
in
a contamination free location.
Both record copies
are considered official
copies.
One is retained
by the licensee
and the other by Westinghouse
upon completion of the refueling.
The licensee
has
no administrative
procedures
governing verification sign offs.
When special
sign off
conditions are involved, the licensee will provide additional instructions
in the particular procedure.
No special
instructions covering sign offs
were contained in the refueling procedures
reviewed by the inspector.
Review of the subject refueling procedures
revealed
the following verifi-
cation sign off discrepancies
and inconsistencies:
a.
Procedure
9.2.6, "Cavity Seal Installation", steps
1 through
12.
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Status:
Work complete
Sign offs made in both record copies;
no sign offs made
in working copy.
b.
Procedure
9.2.6,
"Reactor Cavity Seal
Removal", steps
1 through
4
Status:
Work not started
Sign offs made in one record copy;
no sign offs made in
the second
record
copy or working copy.
c.
,Procedure
9.2.8,
"Reactor
Head Removal", item c.
Status:
Work complete
Sign off not made in either record copy; sign off made
in working copy
d.
Procedure
9.2.8,
"Reactor
Head Installation", steps
A through
E.
(Reactor
Head reinstalled temporarily to support
instrumentation
maintenance
work prior to core alterations.)
Status:
Work complete
No sign offs made in either record copy; sign offs made
in working copy.
'tems
a, c, and
d indicate
a lack of control regarding required verifi-
cation signatures'that
identify satisfactory
work completion.
Item b is
a condition where sign offs were
made for a work operation that had not
commenced.
Failure of the licensee to follow refueling procedures
as
required
by Technical Specification 6.8. 1 is
a violation
(315/85021-02(DRS) ) .
No other violations or deviations
were identified.
5.
Followu
of IE Bulletin 84-03, Refuelin
Cavit
Water Seal
On August 24,
1984, the
NRC issued
IE Bulletin
( IEB) 84-03 to all power
reactor facilities.
The IEB, which described
the events
surrounding
a
refueling cavity water seal failure at the
Haddam
Neck facility, required
licensees
to evaluate
the potential for and consequences
of a seal
failure and to submit
a suranary report supporting their conclusions.
On November 27,
1984, the licensee
submitted the required report.
In
that report the licensee identified design differences
between
the seal
used at D. C.
Cook and the seal
used at Haddam Neck, seal installation
techniques
to be followed, the D. C.
Cook postulated
seal failure accident
based
upon the failure of the inflated portion of the seal,
the capacity
of available cavity water makeup
systems,
an assessment
of no fuel becoming
uncovered,
and emergency
procedures
in place to mitigate the consequences
of such
an event.
During the inspection,
the inspector
reviewed the licensee's
response
and
supporting information which included the potential for loss of refueling
cavity and/or spent fuel pit (SFP) water inventory by means other than
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cavity seal failure with the following results:
a.
The annulus
between
the reactor vessel
and the refueling floor was
measured
to determine
gap dimensions
and inspected for cleanliness
and uniformity.
The nominal
two inch gap
had
a maximum deviation of
0. 109 inch which is in agreement
with original construction
drawings.
b.
c ~
d.
e.
Prior to seal installation,
a 20 psig bubble test
was performed
on
the entire seal ring.
No leaks were detected.
The annulus
was
inspected for sharp
edges,
burrs, etc., that could damage
the seal.
Discontinuities were removed
as required.
Following installation, seal integrity was confirmed by pressurizing
the seal to 45 psig and verifying a pressure
decrease
of less
than
one
psi for one hour.
Seal
pressure
was then
reduced to the operating
range
(15-35 psig)
and the nitrogen supply line relief valve set at
40 psig.
In addition,
RTV was applied to the vessel-seal
and
refueling floor-seal interfaces.
The licensee
is evaluating cavity seal
receipt inspection require-
ments.
Currently, cavity seal
receipt inspection consists of
inspecting for apparent
shipping
damage
and that the item appears
to
be the item ordered.
The licensee is considering
a revision to the
inspection
acceptance
criteria which will require the vendor to
furnish
a "certificate of conformance" to the licensee
assuring that
the Presray seal,
PRS 585, meets material,
dimension,
and hardness
requirements.
The inspector considers
that the upgraded
seal
receipt inspection
criteria are necessary
to assure
seal acceptability.
Revision of
the licensee's
cavity seal
receipt inspection
requirements will be
tracked
as
an open item (50-315/85021-03(DRS)).
A new seal
or a seal
retained
from a previous refueling may be
installed in the refueling cavity.
The seals
have
been
added to the
plant shelf life program to protect from using
a deteriorated
seal.
Present
shelf life has
been established
at 60 months.
The licensee
has
conducted
an evaluation of fuel height drop.
Based
on an evaluation of Westinghouse
and plant drawings,
the maximum
fuel height drop would be
14 inches.
Testing conducted
by TVA and
Duke have demonstrated
seal
adequacy
based
on
a two foot drop onto
the Presray
585 seal installed in a two inch annulus.
Consequently,
the licensee
does not intend to conduct additional tests.
g.
Procedures
are in effect directing that fuel suspended
from either
the Manipulator Crane or Spent
Fuel Pit (SFP)
crane
be placed in a
safe location to prevent
becoming
uncovered
during
a loss of water
accident.
The procedures
also provide instructions for closing the
Transfer Tube Valve and the Weir Gate
on the SFP.
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The licensee
stated that the
SFP could not drain to within two feet
of the stored fuel.
This was
based
on engineering
judgement
following administrative action of aforementioned
procedures
to
recover from a small leak.
It was estimated that should the seal
fail and the
SFP drain to the lowest possible reactor level, seven
inches of water will always cover the fuel.
This is sufficient to
ensure
adequate
cooling.
Fuel in the core will be covered
by
greater levels of water.
i.
The licensee
concludes,
based
on engineering
judgement, that if the
active portion of the seal
(the inflated lower portion)
was to
fail, the passive portion, (the solid wedge
shaped
upper portion) of
the seal
would limit leakage to less
than the makeup capacity of
4500 gallons per hour.
This would provide additional
time for
mitigation.
It is concluded
tha
the licensee
has adequately
resolved the issues
identified in IEB 8 -03 and the
IEB is closed.
During the inspection,
a review was conducted to determine if other
potential
mechanisms for loss of water from the refueling system existed.
These potential
leakage
paths
include NI detector well covers,
sand plug
covers, refueling canal drain covers, refueling cavity floor drain valve,
transfer tube to unit not being refueled,
Residual
Heat
Removal
System,
nozzle
dams
and reactor vessel
head 0-ring seal leakoff
line.
Procedures
are in place to verify that all covers
are properly
sealed
and bolted and that valve line ups are correct prior to flooding.
It was determined that none of these potential
leak paths
would lead to
catastrophic failure resulting in water uncovering stored fuel.
It is concluded that the issue of loss of refueling system water
inventory is adequately
resolved.
No violations or deviations
were identified.
6.
~0
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
3 and 5.d.
7.
Exit Interview
The inspectors
met with licensee
representatives
(denoted
in Paragraph
1)
on
July 18,
1985, to discuss
the scope
and findings of the inspection.
The licensee
acknowledged
the statements
made
by the inspectors
with
respect
to items discussed
in the report.
The inspectors
also discussed
the likely informational content of the inspection report with regard to
documents
or processes
reviewed
by the inspectors
during the inspection.
The licensee
did not identify any such documents/processes
as proprietary.
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