ML17317B453
| ML17317B453 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 06/21/1979 |
| From: | Baker K, Boyd D, Masse R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17317B450 | List: |
| References | |
| 50-315-79-13, 50-316-79-10, NUDOCS 7908150382 | |
| Download: ML17317B453 (18) | |
See also: IR 05000315/1979013
Text
U.S.
NUCLEAR REGULATORY COMMISSION
OFFICE
OF INSPECTION AND ENFORCEMENT
REGION III
Report No. 50-315/79-13;
50-316/79-10
Docket No. 50-315;
50-316
License
No. DPR-58;
Licensee:
American Electric Power Service
Corporation
and Michigan Power
Company
2 Broadway
New York, NY
10004
Facility Name:
Donald C.
Cook Nuclear Power Plant, Units
1 and
2
A
Inspection At:
Donald C.
Cook Site,
Bridgman, Michigan
Inspection Conducted:
Ma
5-31,
1979
Inspectors:
. E.
Masse
g~r
y
K.
. Baker
Approved By:
D.
. Boyd, Acting Chief
Reactor Projects
Section
3
4'~ 7P
l'g~ q
Ins ection
Summar
Ins ection on Ma
5-31
1979
(Re ort No. 50-315/79-13
50-316/79-10)
Safety Features;
assessment
of operating procedures;
review of feedwater
pipe cracking
on Unit
1 and Unit 2; and review of Unit
1 source
rod
location problem.
The inspection involved 230 inspection-hours
by two
NRC inspectors.
Results:
Of the five areas
inspected,
four items of noncompliance
were
noted in two of the areas
(Infraction:
failure to maintain or implement
procedures.
Paragraphs
3.a, 3.d,
and 3.j; and Infraction:
modification
not reviewed or approved
by Onsite Review Committe, Paragraph 3.b).
DETAILS
1.
Persons
Contacted
+D. Shaller, Plant Manager
- B. Svensson,
Assistant Plant Manager
-R. Lease,
Operations
Superintendent
-R. Dudding, Maintenance
Superintendent
+J. Stietzel,
QA Supervisor
-C. Murphy, Production Supervisor, Unit 2
+H. Chadwell, Production Supervisor,
Unit
1
I,. Smith, Shift Operating Engineer
The inspectors
also contacted
a number of operators,
technicians
and
maintenance
personnel
during the course of the inspection.
-Denotes
those present at one or more of the exit
interviews'.
Review of 0 erator Trainin
On April 18,
1979,
an IE and
NRR task group of Messrs.
Streeter
and
Campbell conducted training sessions
for operators
and staff members
responsible for plant operations.
These
sessions
were relative to
the Three Mile Island Nuclear Plant "incident including event chronology,
operator actions,
resultant technical information,
and requirements
being imposed
on licensees
brought about by this incident.
In
addition to these training sessions,
the inspector also discussed
with operators
the following topics (relative to TMI information) to
ascertain
whether knowledge in these
areas
was adequate:
a.
Procedure
changes
brought about by the licensee's
response
to
and 79-06A (Rev. 1).
These bulletins
direct the licensee
to respond to various questions
and require-
ments pending from information received
from the TMI incident.
b.
Operator
knowledge of specific measures
which provide assurance
that Engineered
Safety Features
(ESF) would be available if
required
and, in particular,
measures
for returning such systems
to operable
status
following maintenance
and testing.
c ~
Operator
knowledge of specific and detailed
measures
to assure
that automatic actuations
of ESF are not overridden except
as
directed.
d.
Operator
knowledge of resulting conditions following reset of
ESF for systems affecting control of radioactive
gases
and
liquids.
e.
Operator
and supervisory personnel
knowledge of the provisions
and directives for early notification of serious
events.
Relative to the types of communication
now required,
several
dedicated
phones
were installed
(May 31,
1979) to assist in
open communications
between the plant and
NRC Incident Response
Centers
located in Glen Ellyn, Illinois, (Region III Office)
and Bethesda,
(NRC Headquarters).
Plant phones
are
installed in the following locations:
1.
2.
3.
4.
Unit 1 Control Room
Unit 2 Control Room
Shift Operating Engineer's Office
NRC Resident Inspector's Trailer
No items of noncompliance
or deviations
were identified.
3.
Ins ection of En ineered Safet
Features
(ESF)
The inspectors verified by independent
examinations of records,
procedures
and equipment that
ESF were operable in accordance
with
Technical Specifications
and that procedures
and administrative
controls provided assurance
of continued operability.
The following
paragraphs
relate to those
areas
checked:
a.
The inspectors
reviewed valve lineups, breaker lineups,
and
switch lineups for all ESF systems
and
compared
them with plant
flow diagrams to verify the adequacy of current alignment
procedures.
Numerous inconsistencies
between drawings, proce-
dures,
and as-built status
were identified, concentrated
almost
entirely in the Safety Injection, Residual Heat Removal,
and
Diesel-Generator systems'll
errors
have been
documented
and
passed
along to the licensee for appropriate action.
The inspectors,
using licensee
approved procedures,
physically
verified alignments for ESF components in accessible
areas.
Major 'components
and essential
flow paths
were verified and
a
sampling of minor valves (vents, drains, etc.)
was conducted.
For instances
of operating
components,
observation of operation
was substituted for verification of individual valve positions.
While verifying alignment of Unit 2 diesel-generators
on May 14,
1979, the inspectors
noted that the "after cooler drain valves"
had been
removed from both the
2AB and
2CD diesel generators.
These valves
appear
on flow diagram 12-5150A-14 (J-8).
The
flow diagram indicates that the valves
should be'hut.
Checkoff
sheet 5.1 of procedure
1-OHP 4021.032.006
(Rev. 2) "Emergency
Diesel Generator
Miscellaneous
System Lineup" also requires
the
valves to be shut.
Temporary
Change
No.
1, which had expired
on September
27,
1978, required the valves to be open.
The
inspectors
found that these valves
had been physically removed,
and discussions
with the licensee
indicated that the valves
had
been
removed for an extended period of time due to problems
maintaining them in the open position.
A facility change for
their removal had not been initiated, nor had the
PNSRC reviewed
.
or approved their removal.
These findings represent
apparent
noncompliance with Technical Specification 6.5.1.6.d which
requires
the
PNSRC to review proposed
changes
to equipment
affecting nuclear safety,
and of 10 CFR 50.59 requirements
to
conduct
a safety analysis prior to modification of safety
related
equipment.
The inspector
reviewed administrative controls of the licensee
to assure
that ESF components
or systems
are properly returned
to service following testing or maintenance activities.
The
controls were observed
to be adequate if properly implemented.
The inspector
reviewed
changes
that had been
made to the following
procedures
since the last surveillance tests
had been performed
to verify that,
when testing or maintenance
has
been completed,
the system will be returned to an operable
condition:
1-OHP 4030 STP.002;
1-OHP 4030 STP.004)
1-OHP 4030 STP.005;
1-OHP 4030 STP.007;
1-OHP 4030 STP.008;
1-OHP 4030 STP.009;
1-OHP 4030 STP.010;
1-OHP 4030 STP.Oll;
1"OHP 4030 STP.013)
1-OHP 4030 STP.017;
1"OHP 4030 STP.020;
1-'OHP 4030 STP.022)
1-OHP 4030 STP,024;
1-OHP 4030 STP.025;
1-OHP 4030 STP.026)
1-OHP 4030 STP.027)
1"OHP 4030 STP.033)
1-MHP 4030 STP.005)
1-MHP 4030 STP.006;
1-MHP 4030 STP.008)
1-MHP 4030 STP.009;
1"MHP 4030 STP.010)
1-MHP 4030 STP.011;
1"MHP 4030 STP.012;
1"MHP 4030 STP.013)
1"MHP 4030 STP.014)
2-OHP 4030 STP.002
2"OHP 4030 STP.004
2-OHP 4030 STP.005
2-OHP 4030 STP.007
2-OHP 4030 STP.008
2-OHP 4030 STP.009
2-OHP 4030 STP.010
2-OHP 4030 STP.011
2-OHP 4030 STP.013
2-OHP 4030 STP.017
2-OHP 4030 STP.020
2-OHP 4030 STP.022
2-OHP 4030 STP.024
2-OHP 4030 STP.025
2-OHP 4030 STP.026
2-OHP 4030 STP.027
2-OHP 4030 STP.033
2-MHP 4030 STP.005
2-MHP 4030 STP.006
2-MHP 4030 STP.008
2-MHP 4030 STP.009
2-MHP 4030 STP.010
2-MHP 4030 STP.011
2-MHP 4030 STP.012
2-MHP 4030 STP.013
2-MHP 4030 STP.014
1-THP 4030 STP.Oll;
1-THP 4030 STP.021
1-THP 4030 STP.012;
1-THP 4030 STP.022
1-THP 4030 STP.014;
1-THP 4030 STP.045
1-THP 4030 STP.015;
1-THP 4030 STP.046
1-THP 4030 STP.016;
1-THP 4030 STP.047
1-THP 4030 STP.017;
1-THP 4030 STP.048
1-THP 4030 STP.018;
1-THP 4030 STP.049
1-THP 4030 STP.019;
1-THP 4030 STP.050
1-THP 4030 STP.020;
1-THP 4030 STP.051
1-THP 4030 STP.1007
1-THP 4030 STP.214
1-THP 4030 STP.205)
1-THP 4030 STP.217
2-THP 4030 STP.100)
2-THP 4030 STP.111;
2-THP 4030 STP.112;
2-THP 4030 STP.113;
2-THP 4030 STP.114;
2-THP 4030 STP.119;
2-THP 4030 STP.120)
2-THP 4030 STP.150;
2-THP 4030 STP.205
2-THP 4030 STP.121
2-THP 4030 STP.122
2"THP 4030 STP.145
2"THP 4030 STP.146
2-THP 4030 STP.147
2"THP 4030 STP.148
2-THP 4030 STP.149
2-THP 4030 STP.151
12-THP 4030 STP.203)
12-THP 4030 STP.210
12-THP 4030 STP.204;
12-THP 4030 STP.211
12-THP 4030 STP.207;
12-THP 4030 STP.212
12-THP 4030 STP.208;
12"THP 4030 STP.216
12-THP 4030 STP.209)
12-THP 4030 STP.218
The above current,
approved procedures
include
a return to
status with the following exceptions:
12-THP 4030
1-THP 4030
12-THP 4030
1-THP 4030
1-OHP 4030
2-OHP 4030
1-OHP 4030
2-OHP 4030
STP.203;
STP.205;
STP.207;
STP.214;
STP.002;
STP.002;
STP.004;
STP.004;
12-THP 4030 STP.218
1-OHP 4030 STP.008
2-OHP 4030 STP.008
1"OHP 4030 STP.010
2-OHP 4030 STP.010
1-OHP 4030 STP.005
2-OHP 4030 STP.005
1-OHP 4030 STP.007
2-OHP 4030 STP.007
The inspector also reviewed the following Instrument Maintenance
procedures
for both units:
THP 6030 IMP.014;
THP 6030 IMP.211
THP 6030 IMP.031;
THP 6030 IMP.215
THP 6030 IMP.219;
THP 6030 IMP.226
Exceptions
were noted for:
THP 6030 IMP.211;
THP 6030 IMP.215
THP 6030 IMP.226
Comments
on these procedures
have been
documented
and passed
along to the licensee for appropriate action.
On May 9,
1979, while checking temporary
changes
to procedures,
the inspector noted that procedure
2-OHP 4030 STP.005
"ECCS
Pumps Operability Test" in the Unit 2 Control Room contained
change
No'.
7 dated February 5,
1979.
The temporary
change
was
a carbon
copy with only the first level of review signed off.
A review of plant records
revealed that the temporary
change
had not been
reviewed by the
PNSRC and approved
by the Plant
Manager
as of May 9,
1979.
This finding represents
apparent
noncompliance with Technical Specification 6.8.3.c requiring
review of temporary
changes
to procedures.
The inspector
reviewed the last completed surveillance tests
on
each
ESF system
and verified that acceptance
criteria were met.
The inspector visually observed that pressurizer
level bistables
on Unit 2 had been placed in the trip position as requested
by
NRC and that surveillance
procedures reflect the change.
Unit 1 was shutdown for refueling on April 6,
1979,
and prior
to startup the licensee
intends to implement an
RFC (modification)
which will alleviate the requirement of placing level bistables
in trip.
The change
would allow a two out of three logic on
pressurizer
pressure
channels
only (not coincident with level)
for safeguards
actuation.
This modification will also take
place
on Unit 2 when approved.
The inspector
reviewed administrative controls for returning
ESF systems
to operability following extended
outages.
Plant
operating procedure
1(2)-OHP 4021.001.001
"Plant Heatup From
Cold Shutdown"
and 1(2)-OHP 4021.001.002
"Plant Startup
From
Hot. Standby to Minimum Load" contain administrative
requirements
to return
ESF systems
to operable status.
Additional controls
are provided by the Surveillance Test Program
(PMI-4030 and
implementing procedures),
the Clearance
Permit System
(PMI-2110),
and Job Order Controls
(PMI-2290).
Adherence to these
adminis-
trative controls would appear to be adequate
to assure
systems
are returned to an operable status
following outages.
The inspector
determined that the licensee
does not use independent
verification of valve/breaker/switch
alignments following
outages
and after test or maintenance activities.
The licensee
does
use independent verification for tagging out safety related
equipment to prevent the inadvertent
removal of redundant
equipment
from service.
Independent verification is also used
for radioactive effluent releases
and for jumper/bypass
installation
and removal.
The inspector verified that valves in the Auxiliary Feedwater
System required to be locked and/or sealed
were indeed in the
positions required
and controls in evidence.
A test line
isolation valve was found mispositioned
by the inspector
and
was immediately corrected
by the licensee.
It was brought to
the attention of the licensee that this was another'xample
of
the lack of positive control for which the licensee
was recently
issued
an item of noncompliance.
requirements
for procedure
im
4.
Assessment
of 0 eratin
Procedures
During a diesel generator
alignment check on May 14,
1979, the
inspectors
noted that the "automatic voltage set" for the
2AB
diesel generator
was set at 7.2.
The alignment procedure
(2-OHP 4021.032.006,
Revision 2, with Temporary
Change
No. 2)
required the setting to be 5.7.
A review of records indicated
that the diesel
had been last run on May 10,
1979,
and
was
realigned for automatic start using signoff sheet 5.3 of 2-OHP
4021.032.006,
Revision 2, however,
Temporary
Change
No.
2 had
not been
used to change
the value of the setting from 7.2 to
5,7 on the signoff sheet.
As a result,
the operator set the
value at 7.2.
This action would not have rendered
the diesel
generator
but would have
caused it to operate at
a
slightly higher voltage than desired.
This finding represents
apparent
noncompliance with Technical Specification 6.8.1
plementation.
The inspector verified by observation,
review of procedures,
and
discussion with the licensee
(including operators)
that partial
actuation of Safety Injection (one train) is not used routinely to
assist in pressurizer
level control during routine operational
that in the vent of high pressure
injection, operators
have been instructed to keep
one or more reactor coolant pumps in
operation providing forced flow unless
continued operation would
result in unsafe
or worsening plant conditions; that operators
are
aware of operational criteria for reactor coolant pumps
and have
knowledge of how to identify 50
subcooling in primary loops
as well
0
as saturated
or unsaturated
conditions; that the licensee
does not
have
a procedure for feeding dry steam generators,
however,
the
licensee
did indicate that if all four loops had boiled dry, feed
would be reestablished
to those
whose integrity was
known to be intact.
In the event only one or two steam generators
boiled dry, the licensee
would continue to feed those still in
service but would probably not reestablish
flow to the dry ones.
Tagging procedures
do not address
control of tags to assure that
control panel switches,
indicators,
recorders,
etc.
are not obscured.
This potential
does exist in the Control Rooms
due to board layouts.
This matter has been discussed
with the licensee
and administrative
control in this area
has been developed.
No items of noncompliance
or deviations
were identified.
Feedwater Pi in
Cracks
On May 9,
1979, Unit 2 was brought to a cold shutdown condition to
determine
the source of water accumulating in the containment
sump.
Leakage
was originally suspected
to be from the Nonessential
Service
Water System
(NESW) serving containment ventilation coolers.
Investigation
on May 20,
1979, revealed
small leaks in main feedwater
lines to No.
1 and No.
Cracks
were found adjacent
to the welds connecting
the feedwater piping to, the steam generator
nozzles.
Subsequent
examinations
indicated
cracks also existed
on
all four loops
on both Unit
1 and Unit 2.
The licensee
informed the
Resident Inspector of the initial findings on May 20,
1979.
The
licensee
has initiated an extensive repair program following the
determination of the cause of cracking,
reported to be high cyclic
fatigue failure.
The Resident Inspector
and
a Region III piping
specialist
are following the repair progress.
A separate,
detailed
report will be submitted at
a later date by the piping specialist.
Source
Rod Location Problem
During the Unit
1 refueling,
new secondary
neutron sources
were
installed in locations
90
removed from the original source
rod
0
assemblies.
Following reactor
head
replacement,
the licensee
received
word from Westinghouse
of a potential problem if source
rods were
placed in certain core locations.
The problem would exist in those
locations with swirl vanes
on the upper core plate.
In these locations,
the hub of the source
rod assemblies
would come in contact with the
swirl vane,
and could result in damage to the source rods, fuel
assembly or swirl vanes
due to the excess
weight resting
on the hub.
The licensee
removed the reactor
head
and thoroughly inspected
the
affected areas.
There
was evidence of contact between the hub and
swirl vanes,
however,
no damage
had occurred.
The affected
assembly
and adjacent
assembly
areas
were inspected for bowing, debris, etc.,
and video tapes
were
made of the inspections.
One adjacent
assembly
had
a tom spacer grid tab which was removed.
The sources
were
moved to adjacent locations in the core which did not have swirl
vanes.
The inspector followed associated
activities including a
review of video tapes
through completion of the transfers
and
inspections.
No items of noncompliance
or deviations
were
identified.
8.
Exit Interview
The inspectors
met with licensee
representatives
(denoted in
Paragraph
1) on May ll, 18, 25,
and June
1,
1979, at the conclusion
of weekly inspections.
The inspectors
summarized
the scope
and
findings of the weekly inspections.
Attachment:
Preliminary
Inspection Findings
-9"
>>
~
~
I I
~,
~ ~ I
~
~
4
~
~
~ ~ II
I
>> ~
>>, Z>>>>l Vk
~ ~
~
~
~ ~
~ '
~
>>>>
~
~
~
~'
I '
~
i>>
~
I
~
~ ~
I'
.I
>>
~
~
41 ~ ~
I
~
~
~ ~
~
I
I
,unerxcan Electric. Po<<ex'vc.
Cox'p. '"'.'.;",".
~ Xnd~ana
b 1ffchigan Podex'omp. any
~ '".;:.:
2 j3road<<ay
.
Ne<<York,
NY '10004
)).
C.
Cook Unit 1 'x'~dgrnan,
YX'l
D.
C.
Cool>> Unio 2 '.
oid
ma>>I
YX
2>>. REGXOilAI OI'I'XCE,
~
~
~
I
~
, U.S.. Nuclem
Regu3,atoxJJ
Commiaafon ..
.O~H.ce
oZ Xnopection 6 Enforcement,
1QXX
l&9 Roosevelt
Road
Glen E3.3.yn, XL',603.3/
~
~
~
~
'
~
~
3>>
DOCKET iNR13HRS,
~ ~,.
4..
LXCZilSE YWBERS
5.
DATE OI'NSPECTXOii
50-315
50-316 '
DPR-58 ':DPR-q~:
'
7-
I l
79
2
I
>>
I
~
~
I
~,, ',
~
I
~
CJ
I
6. 'ithin tho scope
o~ the inspect~oa,
no.'terna
og noncomp3.fence
ox',dcv"AtJono ".'-'
.,
~~,
~
~
~
~
~
~
J
',I
I
~
~
'Vo.'ce,o olio<<Xng Rattex'a
A'do p3 el xl~n*MJ..inopecc~on. Ãindingas' l..i."':;,.'I'. '. l':"~,o..<".o
~
Co>>I'I~jj j'b:YEK!<'f4'Ic'II:,4::.Io'<PK'cl'F'Lcl>>IT)Q'i'0:::,:Q>>8:~I c>>$ ;::TiFl,f'cP>'f g':
cl4Q&a
7:::I'g','.Pg'oc i- b'ul"g.";,,-',g @ifP'!QI5$4':;:5+8,,':0Q.>>
Ii ;;G.tie>>P&'a>>e'/,:
'
olio co'o>>,>>>>'Ilo::! gjjQy:.:Pqjj'P'.i:QPe'440'U lQ";:Pi,'c.y::.':PAlh~ "&Pifc'v<gM
EhI
TH<'Pk5'kC::ANC Iform'i2cu'4': ~i:: Ne..."."Y64t':xiii>>;4 eE; t4.le;:tgs~,'..
Co>>>>xmas:
~.O
Ia."O-:'.AO>>lCO',>>-.":4iidCiiiCAriOA'::-:4;gl,'!.":0:;;PE q-via'i4b:.'.l.kO',;
.
.cabu~d
Fo.,'~K:::,,~Sr'r4'it~'H'~ ',8'A'i'H~~j'8'~H"~Q6':-lidi~~ej~ra,
" ...-", ',
'IWC-au~'Sj'::V~jQ<'g~ >i@'~P'j %8','gSP:;:>y<jR+', :,>'liF.'<<N<'~/MATE
L~ 'Ai~T~i++.".'>;.~Hi"'<4N~~ "Fg-tiT; -4 it'ma~'A'gus; L,"IH
~us'8:,~
A~ >"'>t,"C.',,0A, t,'s>>CA'4(N'<~.fjj:b," gv'~i~g>>",FioM':.b'jA':&'/Age'" "'
~
~
>>.
'I
~
~
~
>>
I
~
~ I
~
~
~
~
I
'
>>
~
~
~
I
S>'heoe
pzelkmknax'y"'inspect:Ron
f~nNnpe VX3.1 'be x'evkeve'd by NRC
Supexvf.oion/'"'anat,ement:
at the Reef.on
XXX Od ace an6
~hey vQ.3. coax'espond
vAth you
concexnfhap
AYVJ enfox'cement;
Actions
- i;
- ,-.:,,':;.
'
I
I
p
~
I
~ ~ j
I
~
i
~
~
~
I
I
~ i
~
~
~
~
~
LQJ'iiRY'NSPECTXOH FXNDXNG
I
~
~
P
le
LICENSEE
'merican Electric Power "Svc. Corp.
Xndiana
6 Michigan Power
Company
2 Broadway
.
New Xork, NY
10004
D. C.
Cook Unit 1:
Bridgman,
MX
D. C.
Cook Unit 2
. Bxid man
YX
2.
REGXONAL OFFXCE
U.S. Nuclear Regulatory
Commission
Office of Xnspection
6 Enforcement,
RXXX .
799 Roosevelt
Road
Glen Ellyn, XL'0137
3 '
DOCKET NFiiBERS
50-315
50-316
4.
LXCENSE NUMBERS
5.
DATE OF
XNSPECTXON
'PR-58
'
~
/F
~
~
p
2 oWF )-ozl ozz.~w
4 z.w
Z
uMz
zz ..~
Il
e
- <zi.
/ 6.; Mithin the scope of the inspection,
no items of noncompliance ox'eviations
were found.
~ ~
7.;: The. following matters are prel'iminary".inspection findings
s'ocFgdo.~f
W
FAISKC-W~~~
8/JSA'C..
/~/.8+*
These preliminary'inspection
indings vi11 be'reviewed
by'NRC
Supervision/'anagemerit
at the Region
XXX Office and they Villcorrespond with you
concerning any enforcement
act"'on.
Nuclear Regulatory
Commission Xnspectox
PRELIMINARY XNSPEC'lION FINDINGS
1.
LICENSEE
2.
REGIONAL OFFICE
American Electric Power Svc. Corp.
& Michigan Power
Company
2 Broadway
New York, NY
10004
D. C. Cook Unit 1
(Bridgman, MI)
D. C. Cook Uni
2
(Bridgman, MI)
U.S. Nuclear Regulatory Commission
Office of Inspection
& Enforcement,RIII
799 Roosevelt
Road
Glen Ellyn, XL
60137
3.
DOCKET NUNKS
50-315
50-316
LICENSE NUMBERS
DPR-74
5.
DArE OF INSPECTION
6.
Within the scope of the inspection,
no items of noncompliance or deviations
were found.
8.
These preliminary inspection findings will be reviewed by NRC Supervision/
Management at the Region XII Office and they will correspand with you
concerning
any enforcement action.
Nuclea
Regulatory Commission Xnspector
PRELX
Y INSPECTION FINDINGS
1.
LXCENSEE
2.
RECXONAL OFFXCE
American Elecgric Power Svc.
Coxp.
Xndiana
& Michigan Power
Company
2 Broadway
New York, NY
10004
D.
C.
Cook Unit 1
(Bridgman,
MX)
D. C.
Cook Uni
2
(Bridgman,
MX)
I
3 ~, DOCKET NPiSE.
U.S. Nuclear Regulatory
Commission
Office of Inspection
& Enforcement,RIXX
799 Roosevelt
Road
Glen Ellyn, XL
60137
5.
DYr.E OF INSPECTION
50-315
50-316
DPR>>74
6.
Within the scope of the inspection,
no items of noncompliance or deviations
were found.
7.
The following matters
are preliminary inspection findings:
Q 8.
These preliminary inspection findings will be reviewed by NRC Supervision/
Management at the Region IIX Office and they will correspand with you
concex'ning
any enforcement action.
I
Nucle
Regulatory Commission Xnspector