ML19322A904
| ML19322A904 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 03/12/1973 |
| From: | Jape F, Murphy C, Warnick R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19322A898 | List: |
| References | |
| 50-269-73-01, 50-269-73-1, NUDOCS 7911270660 | |
| Download: ML19322A904 (61) | |
See also: IR 05000269/1973001
Text
_
. .w s
a
.w . .,
. . _
__.. -. . . . . . .
, . _
_
_ ~-
4
.
._
.
.
.
.
t
.
'
%.
e
'
UNITED STATES
(p
,
ATOMIC ENERGY COMMISSION
!
e<
i
DIRECTORATE OF REGULATCRY OPEFATICNS
-
s
1 J
acciou n - suite eis
g+[g~/]f
2m peacarmee sracer, soarwacs,
, , _ , , , , , , , , , , , , , ,
8rans o*
Av un sr 4, cconcia 3o3o 3
BO Inspection Paport No. 50-269/73-1
Licensee : Duke Power Company
Pcwer Building
422 South Church Street
Charlotte, North Carolina
28201
Facility Name: Oconee Unit 1
, ,
Docket No.:
50-269
License No.:
CPPR-33
Category:
B1
,
Location: Oconee County, South Carolina
. ,
Type of License:
Type of Inspection: Routine, Unannounced - December 30, 1972, January 17-19, and
January 23-24, 1973
Special, Announced - January 29 - February 2, 5, and 26,1973
Dates of Inspection: December 30, 1972, January 17-19, 23-24, 29 -
. .u-
February 2, 5 and 26, 1973
'
Dates of Previous Inspection: December 6, 12-14, 16, 18, 20, and 21, 1972
P--incipal Inspector:
R. F. Warnick, Reactor Inspector
Facilities Test and Startup Branch
Accompanying Inspectors:
F. Jape , Reactor Inspector
Facilities Test and Startup Branch
.
M. S. Kidd, Reactor Inspector
Facilities Test and Startup Branch
N. Econemos, Metallurgical Engineer
Facilities Construction Branch
C. M. Campbell, Radiation Specialist
Radiological and Environmental Protection Branch
J. C. Bryant, Senior Inspector, Engineering Section
Facilities Construction Branch
W. D. Kelley, Reactor Inspector
ObD
Facilities Construction Branc.
_.
_ _ _ _ _ _ -
__ __
.
m,__._-.
.
.ua_
. - -.
._ ,w -:k* -
.
- -
.- - - ~ - - -
.
,
_
N
i
.
-
,
-
I
c
3
,.
.
'
'
BD Rpt. No. 50-269/73-1
-2-
N
._
Other Accompanying Personnel:
Dr. P. Doan, Consultant to Directorate
of Licensing - January 18, 1973
Principal Inspector:
ht<
7S
R. 'F. Warnick, Reactor Inspector
Date
Facili les Te t and Startup Branch
Reviewed By: f f x
3 z/73
C. 3. Murphy, geti, @ ief
Wate'
!
Facilities Test and Startup Branch
.
O
!
O
..
'
%
d
e
i
l
.
e_.
u
- --- -
- - -
-. a -
-
m
__
._.
__
_
. . . _ _ -
.
.
.
,
.
N
Rb RPt. No. 50-269/73-1
-3-
.
SUMMARY OF ETNDINGS
I * Enforcement Action
,
.
None
II.
Licensee Action on Previously Identified Enforcement Matters
A.
Violations
1.
Welding Program Deficiencies (See letter to DPC dated
March 8, 1972, Item 5.)
The DPC consultants report on welding deficiencies and
documentation in the DPC welding program was reviewed.
This item is closed.
(See Details VII, paragraph 2.)
B.
Safety Items
None
4
III. New Unresolved Items
73-1/1 Development of Administrative Procedures
1
Licenses personnel are to develop administrative instructions
concerning the use of emergency procedures, alarm proc edures,
maintenance procedures, and operacing procedures prior to
initial core loading.
(See Details II, paragraph 5, and
Details III, paragraphs 2.b.1, 3.c.3, and 3.c.4.)
73-1/2 Resolution of Test Deficiencies
Deficiencies have been identified on five preeperational
tes ts . Three of these require correction before initial
fuel loading and two before initial criticality.
(See
Details II, paragraph 2(A) .)
-
'
73-1/3 Preoperational Tests to be Completed Prior to Initial
[ael Leading
Nine preoperational tests have been identified that are
required to be completed prior to initial fuel loading.
(See Details II, paragraph 2(B)1.)
4
,, s
_
m . .
.
~
=.= - --
- -
- '
- - -
-
.
i
,
'.
'
~
~ %,
. 50-269/73-1
-4-
,
.
RO Rpt. No.
l
73-1/4 Preoperational Tests to be Completed Prior to Initial
Criticality
Fifteen preoperational tests have been identified that are
required to be completed prior to initial criticality.
(See
Details II, paragraph 2(B) 3.)
73-1/5 Analysis and Approval of Preoperational Test Results Recuired
Before Initial Fuel Loading
Fifteen. preoperational tests have been identified that are
required to be reviewed, analyzed, and approved prior to
initial' fuel loading.
(See Details II, paragraph 2 (B)2.)
.l
73-1/6 ' Analysis and Approval of PreopcPational Test Results
Required Prior to Initial Criticality
Eight preoperational test have been identified that are
required to be reviewed, analyzed, and approved prior to
initial criticality.
(See Details II, paragraph 2 (B)4.)
73-1/7 Emergency Operating Procedures
There are seven emerge.ncy conditions for which procedures
will be written ani approved prior to initial criticality
,
and one for which a procedure will be written within three
months.
(Details Il!, paragraph 2.a.)
73-1/8 Alarm Procedures
Alarm procedures for all safety related systems have been
rewritten. These will be approved prior to initial criti-
cality.
(Details III, para',raph 3.b.)
IV.
Status of Previously Reported Unresolved Items
A.
71-7/1 Thin Walled Valves (Also see RO Inspection Report No.
50-269/71-5, Details C.3.)
The licensee's records of the wall thicknesses of valves
within the reactor coolant system pressure boundary were
reviewed at the site and in the Region II office. The following
items remain to be resolved:
1.
Justification for not measuring entire valve body thickness.
2.
Documentatien of error in Ur measurement and its effect on
,
tabulated resalts.
3.
Valve 1-RV-67
.
_.
. . . .
_ ,
.
.-
-
~.:::
-e
E
--o _ m .-
-
\\
__
.
.
.
's
,
,
.
L
,.
/3'
i
30 Rpt. No. 50-269/73-1
-5-
s
-
a.
Calculation of wall thickness according to applicable
codes.
b.
Verification of valve design.
i
(See Details VII, paragraph 3.)
B.
71-10/2 Icop Backflow
The reactor coolant system loop backflow was measured during
the hot functional testing and will be measured again after
initial core loading. This is covered by test procedure
TP-200/12, " Reactor Coolant Pump Flow Test."
This item is
closed.
(See Details I, paragraph 6.)
C.
,71-10/1 Flcwmeter Error Analysis and Tests
BO is awaiting information and action from DPC concerning
this item.
D.
72-8/l Calibration of Licuid Flow Meter on Waste Discharce Line
The calibration of the 0-50 gpm meter on the liquid waste
discharge line has not yet been completed.
(See Details V,
paragraph 3.)
E.
72-8/2 Incomplete Status of Gaseous Effluent Monitors
All of these monitors are new calibrated and fully operational.
This item is closed.
(See Details V, paragraph 4.)
F.
72-8/4 Verification of In-Plant Air Flows
The question of air flow from waste handling areas to clean
areas still exists.
(See Details V, paragraph 5.)
G.
72-8/5 Im tallation of the Solid Waste comcactor
'
Installation of the solid waste compactor has been ecmpleted.
This item is closed.
(See Details V, paragraph 6.)
H.
72-8/7 Spent Resin Transfer Procedure
Draft procedures for testing the resin transfer system
and shipment of resin have been written. This item is
closed.
(See Details V, paragraph 7.)
~
,
- _ .
- _ _ _ - _ _
-
r
.
. ; :_ - _ , w y t. ; r . _ .
.e-
_ . . __
-
nww -
. _ -
.
'
.
,
BO Rpt. No. 50-269/73-1
-6-
.
,
I.
72-8/9 Calibration of Beckman Beta-Mate II
The calibration of the Peta-Mate II has been completed.
.
This item is closed.
(See Details V, paragraph 8.)
J.
72-8/13 Calibration of Process Monitors
Calibration of all process monitors has been completed.
This item is closed.
(See Details V, paragraph 9.)
.
K.
72-8/14 Completion of Decontamination Facilities
Decontamination facilities are tow complete. This item is
closed.
(See Details V, paragraph 10.)
L.
72-8/15 Verification of Hood Air Flows
'
t
The verification of hood air flows has not yet been shown
i
to be satisfactory.
(See Details V, paragraph 11.)
M.
72-8/16 Filter Test of Hydrogen Purge System
2
)
This test was in process. This item is open pending evaluation
of test results.
(See Details V, paragraph 12.)
i
N.
72-11/1 FSAR Training Commitment vs. Actual Training Received
An amendment has been prepare.1 which Ylill revise the FSAR to
agree with training actually received. This. item is closed.
(See Details I, paragraph 3.)
O.
72-11/2 Evaluation of Sample Delivery Line Losses
4
The licensee has agreed to conduct tests to determine sample
i
delivery line losses in'the sampling lines from the unit vent
and the reactor building to their respective process monitors,
after there is sufficient activity in the reactor building to
'
conduct-such tests. This item is closed.
(See Details V,
paragraph 13.)
P.
7J-ll/3 Evaluation of the Efficiency of the Halogen Collection
Media
Laboratory analyses to verify the performance of these units
will be done. This item is closed.
(See Details V, paragraph 14.)
.
.
O
u
-
-
~
e
.
. _ _ _
_g
-.__c
_
__4
_ _ _ . _ _ . -
-
_ _
_ , . . -=
~_a.
_
_
_
.
',
.
'.m
}'
.
'
RO Rpt. No. 50-269/73-1
-7-
Q. '72-11/4 Iodine Condensation Losses
.
Measures have been ' completed to reduce condensation losses
in the outside sample delivery line. This item it: closed.
(See Details V, paragrap a 15.)
R.
72-11/5 Representative Samoling of Liquid Waste During
Discharge of Rad Waste System
The licensee has agreed to obtain a continuous grab sample
during the discharge of the unisolatable low activity waste
tank, in addition to the sample prior to starting the dis'-
charge. The results of the composite sample will be used
to determine the activity discharged. This item is closed.
(See Details V, paragraph 15.)
-
S.
72-11/6 Unit 1 Security Plan
The licensee has defined the Unit 1 security requirements
which have and/or will be implemented prior to initial core
loading.
(See Details I, paragraph 4.)
T.
Axial' Power Imbalance (See RO Inspection Report No. 50-269/70-9,
Summary - Outstanding Item 16a.)
Region II has received a copy of procuiure TP-800/24,
" Power Imbalance Detector Correlation," and has reviewed it.
Comments will be discussed with the licensee on a subsequent
inspection.
U.
Verification of Core Bycass Flew (See RO Insrection Report No.
50-269/70-9, Summary - Outstanding Item 16d.)
The expected core bypass flow has bet n recalculated and updated
to reflect current core and system pre:sure drop information
and as built dimensions. The calculat nns show that the core
,
bypass flow will be 3.67% of the reactor coolant system flow.
This item is closed.
(See Details I, paragraph 5.)
V.
Reactor Coolant Pump Flow Test (See RO Inspection Report No.
50-269/70-9, Sucmarv - Outstanding Item 16r.)
A revised flow coastdown curve has been provided by the licensee.
'the tests have been conducted and the results are in agreement
with the new acceptance criteria. This item is closed.
(See
Details II, paragraph 4.)
,
l
e
. - _
--
- - - -
w-
w . #
e, % w - was,ua
- . A . -
-
__
.
.
l
.
.
i
.
I ),
RQ Rpt. No. 50-269/73-1
-8-
'
)
.
I
W.
Operational Quality Assurance Manual (See letter to DPC dated
May 12, 1972. )
DPC's " Administrative Policy Manual for Operational Quality
Assurance of Nuclear Stations" has been reviewed and all RO
,
comments resolved. This item is closed.
(See Details I,
f
paragraph 7.)
i
V.
Design Changes
A.
Reactor Building Purge Valves
The limitorque valve operators for the two 48 inch butterfly
valves, PR-V1 and PR-V6, the reacter building purge inlet and
outlet valves, have been replaced to provide operators that
are rated for high radiation service. The inspector requested
that the details of this modification, including design review
and approvals, be made available for review at the site during
a subsequent inspection.
i
B.
In-Line Heaters for Nitrogen Supply Line
'
In-line heaters are being installed in the nitrogen supply
line feeding the core flood tanks. The inspector requested
that the details of this modification, including design reviews
{
and approvals, be moue available for review at the site during
a subsequent inspection.
i
C.
Modification to Permit Underwater Removal of the Gate from
the Fuel Transfer Canal 30-Inch Valves
The design e tange makes it possible to remove the valve bonnet
bolts and remove the bonnet and the valve gate for maintenance.
This item is closed.
(See Details I, paragraph 8.)
D.
Core Flood Tank Restrictors
-
A flow restrictor (venturi insert assembly) has been installed
in both of the core flood tank nozzles in the Unit 1 reactor
vessel.
Field construction supervision was provided by B&W while
DpC provided the manpower. Field construction procedures and
applicable specifications used en this project have been identi-
field as those used in the general repair of this vessel. This
item is closed.
(See Details IV, paragraph 5, and Details VII,
paragraph 4.)
.
- - .
.... - -
_ w
_u n
- - , -
- -
-e.,
. - - - -
_
-:
~-
--
-
-
.
%,
'.
.'
l
'
O
BO Rpt. No. 50-269/73-1
-9-
/
-
,
.
\\
N
.-
VI.
Unusual Occurrences
A.
Reactor Coolant Pump Motor Oil Leak and Fire
"
An oil fire was experienced in the Unit 1 reactor building
on and in the in:nediate area of the 1B2 reactor coolant pump.
(Sea Details I, paragraph 9, and attached Advance Inspection
Information Memorandum / dated January 3,1973.)
VII. Other Significant Findings
A.
Steam Generator Tube Leaks
A leaking staam generator tube resulted in extensive non-
destructive testing and failure analysis.
(See Details IV,
'
paragraph 6 and Details VI, paragraph 5.)
VIII. Management Interview
The results of the inspection were discussed with representatives
of Duke Power Company in management interviews held on January 19
by Warnick, Jape, Kidd and Economos, on February 1 by Warnick and
Kidd, and on February 2 by Kelley. The following people were in
attendance:
January 19, 1973
.
Duke Power Comcany (DPC)
J. W. Hampton - Assistant Plant Superintendent
R. M. Koehler - Technical Support Engineer
M. D. McIntosh - Operating Engineer
D. C. Holt - Assistant Nuclear Test Engineer
N. A. Rutherford - Jr. Staff Engineer ,
February 1, 1973
-
.
Duke Power Company (DPC)
J. E. Smith - Plant Superintendent
J. W. Hampton - Assistant Plant Superintendent
R. M. Koehler - Technical Support Engineer
L. E. Summerlin - Staff Engineer
.
-
__ _
.. _ _ -__ _ __ -
_ _ . _
__ __
__-
-- . - . .
x , L , _ _.u
- _ - -
_
,
,
,
,
O
RO,Rpt. No. 50-269/73-1
-10-
February 2, 1973
- Duke Power Company (DPC)
,
D. G. Beam - Project Manager, Construction
D. L. Freeze - Principal Field Engineer
A. R. Hollins - Associate Field Engineer (Welding)
L. R. Davison - Associate Field Engineer (NDP)
.
The following ita
were discussed:
A.
Previously Idencified Enforcement Matters
The status of previously identified enforce. ment matters, as
described in the sumary of findings, was discussed.
.
B.
Previously Reported Unresolved Items
The status of previously reported unresolved items, as described
in the surmary of findings, was discussed.
C.
Change in Inspectors Assigned to Unit 2
The inspector stated that Frank Jape has been assigned as the
Principal Inspector for Unit 2 for the preoperational testing
phase.
R. F. Warnick will continue as the Principal Inspector
for construction items for Unit 2.
Warnick will continue as
principal inspector for Unit 1 and E. J. Vallish will continue
as principal inspector for Unit 3,
!
D.
Lifted Leads
The inspector stated that as a followup to previous discussions
concerning lifted leads, Constructien's termination check of the
engineered safeguards and reactor protactive system panels for
lifted leads was reviewed and there are no furthur questions.
(See Details I, paragraph 10.)
-
'
.
E.
' Design Changes
DPC was informed that the venturi insert assembly modification
to the reactor vessel core flood nozzles and the modification of
the fuel transfer canal 30-inch gate valves had been reviewed and
the RO inspectors had no further comment at this time.
(See
Details I, paragraph 8, and Details VII, paragraph 4.)
_ _
-
y
I
1
.w a
_ _ _ . _ _ _ . .
_ . ,
c . m.
- -
m
m-
u
- - - - - - -
,
.
.
'
q
'.
RO Rpt.No. 50-269/73-1
-11-'
i
!
The inspector confirmed his request to see the design change
documentation for the installation of in-line heaters for the
core flood tank nitrogen supply line and the replacement of
valve operators on two engineered safeguard valves, PR-V1
and PR-V6.
(See Details I, paragraph 8.)
F.
Fuel Loading and Initial Criticality
The status of items remaining to be completed before fuel
loading and initial criticality was discussed.
(See Details I,
paragraph 11.)
i
G.
Status of Precperational Testing Program
The status of the preoperational testing program was reviewed.
Questions and comments on several tests were discussed and
tests were reviewed with the licensee's representative.
(See
Details II, paragraph 2.)
H.
Review of SRC and NSRC Activities
The inspector stated that he had reviewed the minutes of the
Station Review Committee (SRC) and the Nuclear Safety Review
Comittee (NSRC) .
(Sea Details II, paragraph 3.)
I.
Review of Operating and Maintenance Procedures
9
F. Jape stated that he had completed his review of operating
and maintenance procedures.
'
It appears that the administrative procedures currently in
preparation will resolve all previous questions and ec=ments
on the operating and maintenance proceduren.
(see Details II,
paragraph 5.)
J.
Fuel Inspection
~
An audit of the licensee's inspection of fuel assemblies was
conducted. The inspection records were examined and the in-
spection method was discussed with the licensee's represen-
{
tative. The inspector had no coment on the inspection program.
(See Details II, paragraph 6.)
K.
Administrative Instructions
The inspector st1ted that Administrative instructions regarding
the use of emergency and alarm procedures should be written.
.
_ _ . _ _ _ . -
.w=
~
- - - -
-
- - -
-
--
-
--
.
w
,
.
.
' q
,
RO Rpt. No. 50-169/73-1
-12-
.
,
- - _ _
A member of licensee management stated that these would be
written and approved prior to initial core loading.
(See
,
Details III, paragraphs 2.b.1, 3.c.3, and 3.c.4.)
L.
Emergency Procedures
1.
Emergency Procedures Needed
l
i
The inspector stated that review of emergency procedures
revealed that eight conditions were not yet covered. He
'
also stated that it was his understanding that procedures
i
for seven of these would be written and approved prior to
initial criticality and that the other one would be written
and approved within three months.
'
.
Licensee management stated that this understanding was
correct.
(See Details III, paragraph 2.a.)
2.
Comments on Individual Procedures
The inspector stated that specific comments on emergency
procedures had been given to station persennel an-1 that
he had no questions regarding DPC's plans to resolve the
Comme 9ts.
Licensee management stated that the coments would be
incorporated into the procedures within the time periods
which had been agreed to.
(See Details III, paragraph 2.b.)
M.
Alarm Procedures
The inspector stated that he had reviewed ten percent of the
alarm procedures for Unit 1 and except for comments already
-
given oa selected ones, they were compatible with the guide-
lines cf ANS 3.2.
He stated that these safety related alarm
-
proced. *as which had been rewritten should be approved prior
to initial criticality.
Licensee management stated that ccmments on individual
,
procedures would be incorporated within the time perieds
,
agreed to and that all safety related alarm procedures would
be appaved prior to initial criticality.
(See Details III,
paragraphs 3.b and 3.c.)
.
...
_
,_l
_ _ _ _ _ _ .
-
__
.
_ _ .
___., _ .
-
_ . , _ -
I
.
,
',
ll
.
A
i
RO.Rpt. No. 50-269/73-1
-13-
,
N.
Periodid- Test Procedures and Instrument Procedures
.
The inspector stated that lie had reviewed thirty periodic
test and instrument procedures for Unit 1 and found them
to be quite compatible with the guidelines of ANS 3.2.
He also stated that there were two procedures which he had
consnents on and that agreement had been reached with station
~
personnel regarding CPC plans to take corrective action on
those comments.
Licensee management indicated that the comments would be
resolved.
(See Details III, paragraph 4.)
.
O.
Periodic Test Schedule
.
The inspector stated that he had reviewed the tentative plans
which the operations and technical support groups had developed
for assuring that testing required by Technical Specifications
is completed in a timely manner and that he had no comments
concerning those plans.
He stated that these plans should be finalized and a detailed
testing schedule drawn up very soon since it is required to be
established as soon as a license is received.
3
Licensee management indicated t. hat the ' schedule would be finalized
as soon as possible.
(See Details III, paragraph 5.)
.
P.
In-service Baseline Inspection
The inspector noted that he had reviewed the final in-service
inspection report prepared and submitted by B&W cn February 5
and, based on this review, it appears the report was in compliance
with the " inservice inspection philosophy" of section XI of the
ASME code, insofar as establishing a baseline parameter to be
-
used in comparing results of subseque t Auservice examinations.
The inspector noted that significant reportable indications
'
(defects) were detectied in three areas.. These were, the transi-
tion ring between the once thru steam generator (CTSG) lower head
and the support skirt, the IB CrrSG reactor coolant outlet nozzles,
and the spray nozzle on the pressurizer. In reference to the
support skirt, management reported that B&W performed an
engineering evaluation using a fracture mechanics approach in
l
order to resolve whether these defects impair the structural
adequacy of the vessel or whether the defects are insignificant
.
?
__w
-
. .. a . n + -
--
a.' -
.s--~~--
- ~ ^~
~
" ^ ^
^ '
.
.
.
.
O
RO Rtp. No. S'0-269/73-1
-14-
I
.
and can be safely ignored. Management stated that based on
results of this analysis, it was their opinion that the defects
as indicated by the base-line ultrasonic inspection would not
in any way jeopardize the vessel from performing as designed.
In reference to both nozzles on the IB CTSG and the spray
nozzle on the pressurizer, management reported that areas with
indications exceeding tyr acceptance standards established for
the baseline examination were re-evaluated by B&W and Duke
personnel using the original shop radiographs, and the welds
were found to be acceptable. Hence, these components were
acc2ptable' and repair was not required.
(See Details IV,
paragraph 2.)
Q.
Reactor Coolant System Oil Fire
The inspector noted that heavy deposits of baked-on oil'
residues were detected on the pipe and core coolant pump
housing and suggested that these sur. faces should be cleaned
j
sufficiently to meet the cleanliness requirements of the
i
applicable specification. Management stated that work had
begun on these parts, the objective being to restore the
affected surfaces to their original condition as much as
j
possible.
(See Details IV, paragraph 3.)
'
R.
Reactor Internals,
~'*
The inspector noted that he had conducted a spot check of
the pressure vessel and the internals. This inspection
prod' iced no evidence of excessive wear as a result of the
recently completed hot functional tests.
The inspector also noted that he had conducted a epxprehensive
review of the cleaniness records for the internals in Unit 1
and had no further comments at this time.
(See Details IV,
paragraph 4.)
-
S.
Core Flood Line Restrictor
The inspector noted that he reviewed the documents controlling
field construction of the flow restrictor. He found them to be
in accordance with the applicable ASME codes therefore had
no further comments at this time.
(See Details IV, paragraph 5.)
,
'
.
,, -.
.
.-
.-- -
.
- -
u. . . : .
.. -
--
-
--
w-
. . . _ . , - -
-
- _ _
.
w
,
.
.
,
RO Rpt. No. 50-269/73-1
-15-
i
.
N
T.
Steam Generator Tube Examinations
The inspector noted that Eddy Current (EC) examination had been
performed on 871 of tubes in steam generators lA and IB, with
preliminary results indicating nothing that had not been
previously identified. The inspector also stated that Region II
wished to review the final report of the metallurgical investi-
gation conducted on the tubes which exhibited "intergranular
corrosion (cracking)". Management indicated that a copy of
the aforementioned report would be made available when issued
by Babcock and Wilcox (B&W) .
(See Details IV, paragraph 6.)
U.
Hanger Response and Pipe Vibration
.
The principal inspector stated that the results of hanger
responses and pipe vibration tests were reviewed. In TP-600/14,
" Pipe and Component Hanger Hot Deflection and Inspection Test,"
the inspector determined that the acceptance criteria for RCP-1A2,
at 450' and 500*F in the X direction, had not been met.
DPC completed an engineering evaluation of the excessive deflection,
prior to the end of the inspection, which concluded that the
allowable stress values had not been exceeded.
(See Details VI,
paragraph 3.)
.
._
i
-
'
l
l
.
f
.-
.i
- _.
- ~ . - - .
.
._
+
!
,
.
'
.
.
,
.
.
t
Duke Power Company
f
Oconee 1
RO Eeport No. 50-269/73-1
MAR 151973
P
.
ec w/ encl:
J. G. Keppler, RO
J. B. Henderson, R0
B0:HQ (5)
DR Central Files
Regulatory Standards (3)
Directorate of Licensing (13)
cc encl. only:
4
'PDR
' Local PDR
'NSIC
'DTIE, OR
' State
'To be dispatched at a later date.
.
m
.
.- ~ _ _
__
-._.
__ _ .
. _ -
_ - .. . . - . _ _ . . . _ _
..
_ .- . . . . _ _ _ . _ _ . _ .
. . _ _ . _
_ - . .
. ..
.-_._ ..
..- _
6
-
-
.
i
s
'
j'.
.
.
-
.
f
t.
rS , ,.
.
,
Ltr to Duke Power Company
i-
.
_'
ata MAR 151973
i
.
4
ec w/ encl:
.
i
J. G. Keppler, RO
-
.
J. B. Henderson, RO
,
i
B0:HQ (h)
Directorate of Licensing (h)
M h DR Central Files
J
,
.
ec w/o enc 1:
Iccal PDR
MH, 2
'
i
State
1
i
!
t
i
1
i
l
9
,
i
l
i
e
i
e
i
k
.
$
-
'
i
1.
-
-
l
!
!
-. . . - -
.
.
.
.,
i
2
e
- -.
--
-e..
. - . - . , .. . - , . .
-. ,,,
, . , . . , .
.-
-._,v.-..,,-..,
, ,
..
,