ML17319B045
| ML17319B045 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 08/27/1981 |
| From: | Danielson D, Yin I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17319B043 | List: |
| References | |
| 50-315-81-19, 50-316-81-22, IEB-79-14, NUDOCS 8109150301 | |
| Download: ML17319B045 (18) | |
See also: IR 05000315/1981019
Text
U.S.
NUCLEAR REGULATOR COMMISSION
OFFICE
OF INSPECTION AND ENFORCEMENT
REGION III
Reports
No. 50-315/81-19;
50-316/81-22
Docket Nos. 50-315; 50-316
Licenses
No. DPR-58;
Licensee:
American Electric Power Service Corporation
and Michigan Power
Company
2 Broadway
New York, NY
10004
Facility Name:
D. C.
Cook Nuclear, Plant,, Units
1 and
2
Inspection At:
American Electric Power Service Corporation,
New York, NY
Inspection
Conducted:
August 10-12,
1981
Inspector:
I. T. Yin
Z a7Z/
Approved By:
D. H.=Danielson,
Chief
Materials and Processes
Section
Ins ection
Summar
Ins ection on Au ust 10-12
1981
(Re orts No. 50-315/81-19
50-316/81-22)
Areas Ins ected:
Licensee actions relative to IE Bulletin No. 79-14; re-
view of LER corrective actions.
The inspection involved 23 inspector-hours
at the corporate
engineers office by one
NRC inspector.
Results:
Of the areas
inspected,
two apparent violations were identified.
~pailure to follow design calculation verification procedures
paragraph
6.c.(1); failure to report events
as required by Technical Specifications-
Paragraphs
6.a.(4)
and 6.d.(l)).
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Persons
Contacted
DETAILS
American Electric Power Service
Cor oration
(AEP)
"T.
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+K.
~'G.
"R
'S.
"T
'>S.
~H.
"H
J.
R. Satyan-Sharma,
Nuclear Safety
and Licensing
I. Castresana,
Safety and Licensing Section Manager
H. Vehstedt, Nuclear Safety and Licensing
P. Field,
AVP and Chief Design Engineer
F. Kroeger,
Manager of (}A
H. Steinhart,
Mechanical Engineering
W. Baker, Mechanical Engineering
PSV Section Manager
Sun, Mechanical Design
Ulan, Mechanical Design
Sobol, Nuclear Engineering
T. Sinclair, Design Staff
L. Williams, Section Manager,
Mechanical Design
"-Denotes those
who attended
the exit interview.
Functional or Pro
ram Areas Ins ected
Status of S stem Evaluation
and Modifications (IE Bulletin No. 79-14)
Engineering evaluation for all safety related piping systems in both
Units
1 and
2 is completed.
Hardware modifications had been completed based
on the results of
393 computer calculations.
At the time of this visit, there were 24
computer calculations
which required site modification.
The original Class
1 piping systems
design
and analysis responsibili-
ties were as follows:
D~esi n
Seismic Anal sis
Primary Coolant Loop
MS from SG to Containment Penetration
FW from Containment Penetration
to,SG
All other Class
1 Piping
SSL
AEP
AEP
AEP
W
SSL
SSL
EDS,S.F.
The
was
IE Bulle'tin No. 79-14 computer Class
1 piping systems
analysis
work
assigned
as follows:
EDS,NY-33 calculations
Associated Technologies,
Inc. (ATI) - 32 calculations
Harstead - 4 calculations
Teledyne Engineering Service
(TES) - 2 calculations
AEP - 13 calculations
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2.
Review of Licensee Evaluation Criteria
a.
The original piping design loading conditions
and stress limits
documented in CESAR Volume IV, Appendix B, Table
2 stated:
Upset Conditions:
(a)
Pm < 1.2S
(including OBE)
(b)
Pm (or Pl) + Pb< 1.2S
Emergency Conditions:
(a)
Pm < 1.2S
(Including DBE)
(b)
Pm (or Pl) + Pb< 1.8S
Where:
Pm = primary general
membrane stress
Pl = primary local membrane
stress
Pb = primary bending stress
S = allowable stress
from USASI B31.1,
Code for Pressure
Piping
In discussion with the licensee, it was stated that the original
calculation methods
employed were as follows:
Large Bore Piping Systems:
Class
1 - Computer analysis
(2
~pl>
and larger)
Class
2 - Both computer analysis
and static coefficient
Class
3 - Mostly static coefficient
analysis
Small Bore Piping Systems:
All classes
were based
on static
(2" and smaller)
coefficient analysis
The computer programs
used during the original design were the
programs established
by W, S&L, and
EDS.
The static coefficient
analysis
was based
on EDS Procedure,
"Alternative Piping Analysis
Criteria for Earthquake
and Gravity Ioads for Donald C.
Cook
Nuclear Plant," dated
September
1971.
computer seismic analyses
were performed by EDS, ATI, AEP and other
companies
using their own computer programs.
The static analysis
is still based
on the September
1971
EDS procedure.
b.
The D.
C.
Cook FSAR does not address
piping component
support design
and analysis criteria.
The design requirements
are addressed
in AEP
Specification
No.
DCCPM402QCN, "Pipe Supports,
Special Requirements
for Seismic Class I, II, and Quality Ievel
5 Systems,"
Revision 2,
dated April 20,
1970.
The IE Bulletin No. 79-14 hanger design
evaluations
are also based
on this specification.
In discussions
with the
AEP engineering staff, it was,stated
that
the system discrepancy evaluation operability criteria under the
DBE conditions is the
same
as the original Code stress criteria.
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No items of noncompliance
or deviations
were identified as
a
result of the review.
The EDS static coefficient analysis
procedure,
and the
AEP pipe hanger design specification will
be reviewed in detail during
a future inspection.
3.
Review of En ineerin
Evaluation Procedures
The following AEP evaluation procedures
requirements
were either revised or added to the program since the
inspector's last visit in September
1979
(Region III Inspection
Reports
No. 50-315/79-22;
50-316/79-19):
79-14-1-S, "Structural Design Procedure,"
Revision 1, dated
October
16,
1979.
79-14-4,
"Method of Document Comparison to Evaluate Exiting
Erected Configurations," Revision 2, dated October
16,
1979.
79-1'4-5,
"Procedure for Seismic Evaluation of Discrepancy Data,"
Revision 4, dated January 6,
1981.
79-14-7,
"Procedure for Up-Dating Drawings and Documents Relating
to IE Bulletin No. 79-14,",Revision 2, dated November
17,
1980.
79-14-8, "Close-out Procedure
for. Handling Storage
and Reproduction
of Documentation for NRC IE Bulletin No. 79-14", Revision 1, dated
October 31,
1981.
The inspector
reviewed these procedures.
No items of noncompliance
or
deviations
were identified.
4.
Review of Field Xns ection Walkdown Procedures
During the site inspection performed
on August 29,
1979
(Region XII
Inspection Reports
No. 50-315/79-20;
50-316/79-17)
the inspector re-
viewed Procedure
No.
12-gHP-SP.002,
"Seismic Class I Piping System
Field Inspection for Conformity with Design Documentation of .Seismic
Analyses," Revision 0, dated August 6,
1979,
and had several
comments.
During this visit, the inspector
reviewed Revision
1 of the procedure
dated August 29,
1979, together with a letter from AEP to Region IIX,
AEP:NRC:
00238B,
dated
September
14,
1979,
where
a description of the
field inspection methods being employed is given.
The inspector
con-
siders
the intent of the subject matter has
been met,
and stated that
he has
no further questions.
5.
Followu
Resident
Ins ector (RI) Identified Problems
On April 15,
1981, the Region XII RI while observing the SI and blackout
surveillance,
heard unusual vibration from Support
2GESW-R38 of the Diesel
Enginer Aftercooler System in the Unit 2 diesel generator
area.
This and
other supports in the area
were found loose,
some not supporting the
pipinq system.
He also found that the Unit 2 supports differ from those
in Uni;t l.
A number of hanger detail sheets
were furnished by the RI for
4-
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the inspector's
followup review and evaluation in conjunction with the
review of licensee's
implementation of IE Bulletin No. 79-14 requirements.
The following hangers in question
were reviewed by the inspector:
Calculation 2-006 - Diesel Generator
(DG) 2AB and
2CD Iso.
No.
2-GESW-34 and
The hangers in
question
were R41, R40, R39,
and R38.
Calculation 2-005 - DG 2AB and
2CD Iso. No.
2-GESW-36
and
2-GESW-38.
The hangers in question
were
R44,
V45, V42, and V43.
Calculation 1-005 - DG lAB and
1CD Iso.
No.
1-GESW-35
and
1-GESW-37.
The hangers in question were
R14, R15, R12,
and R13.
Calculation 1-006 - DG lAB and
1CD Iso. No.
1-GESW-33
and
1-GESW-41.
The hangers in question were
R14, R15, R12,
and R13.
The problems identified by the RI included:
(1) hanger locations
were
changed
from pipe runs to pipe elbows,
(2) where variable spring hangers
were called for on design details, rigid hangers
were installed,
and (3)
during the IE Bulletin No. 79-14 walkdown hanger deficiencies,
such
as
bend rod, missing U-bolts,
and clearances
between trapeze
type pipe
supports
and the bottom of the pipe were apparently not being identified
by the site inspection staff.
Relative to problems
(1) and (2) above,
the inspector
reviewed
some of
the site inspection
and corporate engineering evaluation packages
and
found that discrepancies
on V45, V43, R44,
and V42, R38, F39, F40,
and
R41 had been identified during the September
28,
1979 system walkdown;
and the discrepancies
on R8, R9, V10, and Vll had been identified dur-
ing the August 29,
1979
and November 8,
1979 walkdown inspections.
The
inspector
reviewed the stress
analysis
'packages
and concurred with the
licensee that the existing arrangement
does not present
any safety
problem.
This was based
on the fact that the system temperature
is
below 100 F, there was,a high system natural frequency
(some
exceeded
0
33 cps at first mode),
and both primary and secondary
stresses
including
OBE and
DBE are very small in comparison with the
Code allowables.
Relative to (3) above,
the missing U-bolt was determined to be
a mis-
interpretation deficiency on the part of the field walkdown group and
did not represent
a safety problem.
However,
the problems involving
the
damaged
component, i.e., bent hanger
rods
and improper hanger 'in-
stallation, including gaps between supports
and pipes will be followed
up by the inspector during
a future site inspection to assess
whether
or not these
are isolated
cases
or a generic problem.
This is con-
sidered
an unresolved
item (315/81-19-01;
316/81-22-01).
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6.
Review of IE Bulletin No. 79-14 Evaluation Packa
es
A number of the subject packages
were selected
by the inspector for
review.
The review findings are
as follows:
a.
EDS Calculation No. 2-215,
"Chemical and Volume Control,
1-CCS-5,
3" and 4" CS Line From CPN-37," Revision 1, dated October
18,
1980.
(1)
The inspector
reviewed the
EDS Nuclear Inc. Project Instruc-
tions,
"D. C.
Cook Nuclear Plant - NRC Bulletin No. 79-14
Reanalysis",
Revision 1, dated August 13,
1980,
and had no
adverse
comment.
(2)
Valves
QCM-250 and SV-50, the weights modeled in the computer,
were based
on values
from the manufacture
drawing.
(3)
Maximum DBE XY Quake Stress
of 28,749 psi at
C16 pipe elbow
was 99.8'/ of Code allowable (28,800,psi).
In view the fact
that the
SV discharge thrust loads
and the loads
due to the
attached non-safety related discharge
pipe section
had not
been included in the calculation,
the adequacy of design in
terms of meeting the
Code requirements
remained to be ques-
tionable.
Insufficient evaluation of SV discharge piping loads
and the
dynamic effect of secondary piping is,a problem that has been
identified at several facilities.
This matter will be for-
warded to IE Headquarters
for evaluation
as
a potentially
generic problem,
(4)
In review of Revision
0 of the calculation,
dated October
14,
1980, the following stresses
were found:
XY OBE 21,924 psi, which is 114.2g, of Code allowable
(19,200 psi) at C22 pipe elbow, Joint No. 77.
XY DBE 42,515 psi, which is 147.6/ of Code allowable
(28,800 psi) at C22 pipe elbow, Joint No. 77.
ZY DBE 34,342 psi, which 119.2/ of Code allowable
(28,800 psi) at C09 elbow, Joint No.. 23.
In order to correct these overstress
conditions
an X-stop
was added at Joint No. 52A (called 2-GCS-R-608A)
and support
2-BCS-R-610 at Joint No.
74 was modified from a Y-stop to
a
.Y-stop and
a lateral.
The design
was forwarded to the site
on March 6,
1981,
and installation was completed
on April 10,
1981.
An IER was not issued informing Region III that the
piping was overstressed.
b.
ATI Calculation 1-220,
"10" Line From Accumulators
No.
2 and
No.
3 to CPS-17
and
CPS-21 (1-CSI-3)", Revision 0.
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(1)
Check Valve 51CN08XX, the weight modeled in the computer
was based
on valves
from the manufacturer
drawing.
The water
weight inside the valve was included in the calculation.
(2)
Pipe penetration Fl. El. was marked 601'-3/8" but in fact,
it should be Fl. EL. 608'",
However, this did not affect
the computer input.
In discussion with the President of ATI,
it was stated that computer input deck was verified prior to
run,
and
some outputs
were graphically plotted by computer.
The inspector noted that the verification measures
did not
appear to be applied systematically.
A more detailed procedural
review relative to computer
calculation 'verification methods will be conducted by the
inspector during
a future inspection.
This is an unresolved
item (315/81-19-02;
316/81-22-02)
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c.
AEP Calculation
2 -258,
"From Pump 2-RC-20 to Excess
Ietdown Heat
Exchanger,
2-CS-670",
dated
December
3,
1979.
(1)
There
was
no documented
evidence that the computer calculation
was verified by a checker.
(2)
X-restraint
was added to the system to reduce
the seismic stress
to meet the
Code allowable valve.
A
hanger drawing was issued
on December
14,
1979,
and sent to
the site on .December
15,
1979.
The restraint
was installed
on December
21,
1979.
A Licensee
Event Report
(LER) was
forwarded to Region III.
d.
TES Calculation 2-178, "Technical Report,
TR-4128-2, Stress
Analysis of Essential
Service Water to Axuiliary Feed
Pump
Piping (2-ESW-5) for D.
C.
Cook Nuclear Power Plant," dated
January
8,
1980.
(1)
In Paragraph
2 it stated,
"Conclusion, that Analysis of the
system as-built condition showed failure at Nodes
53 through
57.
To correct this condition,
a
Z direction (rigid) re-
straint was added at Node 48 and an X direction (rigid)
restraint
was
added at Node 5013."
Modified 2GESW-R-120
(the Z-restraint) - The drawing
was sent to the site on January
24,
1980.
Added 2AESW-I,-120A (the X-restraint) - The drawing was
sent to the site on January
17,
1980.
The installation of these restraints
was completed
on
February 25,
1980.
No LER was sent to Region III.
Subsequent
to the review, the inspector
determined:
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Item 6.c.(1) are in noncompliance with the
AEP Design Division
Organization
and Procedures
Manual, General Design Procedure
No. 8,
Revision
2 dated October 31, '1980, "Calculations," "Computer Calcula-
tions", Paragraph
2 states,
"The engineer or designer
responsible for
the problem shall stamp the first sheet of output containing the input
data
and review it for accuracy, against the input data
and review or
for accuracy against the input data. 'he'assigned
"Checker" shall" also
check the input data in the
same manner.
After checking,
the originator
and checker shall sign and date the stamp."
This is an apparent viola-
tion identified in Appendix A (315/81-19-03;
316/81-22-03).
Items 6.a.(4)
and 6.d.(1) were in noncompliance with the Technical Specification 6.9.1.8.
The D.
C.
Cook Nuclear Plant, Units
1 and 2,
Technical Specification,
Amendment No. 23, "Prompt Notification with
Written Followup" Section 6.9.1.8 states
that,
"The types of events
listed below shall be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone
and
confirmed by telegraph,
mailgram, or facsimile transmission
to the
Director of the Regional Office, or his designate
no later than the
first working day following the event, with a written followup report
within 14 days.
The written followup report shall include,
as
a
minimum,
a completed
copy of a licensee
event report form.
Information
provided on the licensee
event report form shall be supplemented,
as
needed,
by additional narrative material to provide complete explanation
of the circumstances
surrounding the event."
Section 6.9.1.8.i states,
"Performance of structures,
systems,
or components
that, require remedial
action or corrective measures
to prevent operation in a manner less
conservative
than assumed in the accident analyses in the safety
analysis report or technical specification bases;
or discovery during
unit life of conditions not specifically considered in the safety
analysis report or technical specifications that require- remedial
action or corrective measures
to prevent the existence
or development
of an unsafe condition."
During the review, the inspector
was presented
the following LERs
relative to problems identified as
a result of the IE Bulletin
No. 79-14 evaluations;
however,
none of them described
the above
problem areas:
LER No. R079-050/OIT-O, reported
on January
11,
1980
IER No. R079-065/OIT-O,
reported
on January
17,
1980
IER No. R079=049/OIT-O,
reported
on December
21,
1979
This is an apparent violation identified in Appendix A (315/81-19-4;
316/81-22-04).
Pum
(RCP) Restraint
Problem
The licensee reported to Region III on July 29,
1981 that the shim pack
retaining bolts
on the
RCP No.
12 (Unit 1,
Pump No. 2) were sheared off,
allowing the shim pack to fall from its installed location.
The re-
straint was located at RCP support,
column No.
2 away from the reactor
vessel.
The shim pack (with Gap
C) was designed
to receive
loa'ds
normal to the cold leg during LOCA conditions.
Licensee
engineers
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stated that the slightly offset gap
was based
on pipe thermal expansion
and contraction
movements
and (1) blackout condition of 620 F, (2) hot
functional of 547 F, (3) hot leg reactor full load at 599 F, and (4)
cold leg reactor full load of 540 F.
Subsequent
to the discussion with
the
AEP nuclear licensing
and design staff, the inspector determined:
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AEP engineers
stated that the failure was considered
to be in-
cidental because
the previous
60 plus system transient
cycles
did not have similar problems.
They stated that site personnel
had inspected all other restraint settings at the
and the
and found no abnormal conditions.
A subsequent
review of IER RO 81-026/99T-0 noted that it did not mention the
gaps being checked.
This is an unresolved item
(315/81-19-05;
316/81-22-05).
b.
The original Westinghouse
(Q) restraint
gap setting at hot func-
tional temperature
of 547
F was 0<'nd 1/32".
The AEP design
department
revised the settings to 1/16" with a tolerance of +1/32"
and -0".
Documentation indicating
W engineering
had concurred with
this revised setting
was not available.
The licensee
stated
they
would obtain the necessary
documentation.
This is an unresolved
item (315/81-19-06;
316/81-22-06).
Unresolved Items
Unresolved items are matters
about which. more information is required in
order to ascertain
whether they are a'cceptable
items,
items of noncompliance,
or deviations.
Three unresolved
items disclosed
during the inspection are
discussed in Paragraphs
5, 6.b.(2),
7.a
and 7.b.
Exit Interview
The inspector
met with licensee
representatives
prior to the conclusion of
the inspection.
The inspector
summarized
the scope
and findings of the
inspection.
The licensee
acknowledged
the findings reported herein.
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