ML15224A620
| ML15224A620 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 01/10/1990 |
| From: | Blake J, Coley J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML15224A618 | List: |
| References | |
| 50-269-89-37, 50-270-89-37, 50-287-89-37, NUDOCS 9002020232 | |
| Download: ML15224A620 (8) | |
See also: IR 05000269/1989037
Text
% REGL4
UNITED STATES
o
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W.
V ,t :~ATLANTA, GEORGIA 30323
Report Nos.:
50-269/8.9-37, 50-270/89-37, and 50-287/89-37
Licensee:
Duke Power Company
422 South Church Street
Charlotte, NC 28242
Docket Nos.: 50-269, 50-270,
and
and 50-287
Facility Name:
Oconee 1, 2, and 3
Inspection C
e
November 27 -
December 1, 1989
inspecto :
-
/0 /o
%-'-.
oley
Date Signed
Approves by
4 ;AI I//
4/ $
.
3
lake, Section Chief
bate Signed
M t
als and Processes Section
ng neering Branch
Division of Reactor Safety
SUMMARY
Scope:
This routine, unannounced inspection was in the areas of inservice inspection
(ISI)
and included observation of work and work activities, and review of
licensee's actions regarding
a Unit 3,
2nd Interval,
request for relief
(No. 89-08) on 3A2 and 3B1 reactor coolant pump flange stud holes.
Results:
During the inspector's observation of ultrasonic examination work, two areas of
concern were identified. Violation 50-287/89-37-02, paragraph 2.a., documented
that the 2nd interval inservice inspection plan has not been properly reviewed
in that:
calibration block number 40367 designated for Item Number B05.050.012
Weld Number 3PSP-1 did not have a surface finish that represented the pipe to
be examined (surface cracks and material separations were observed in the
calibration block).
The calibration block also was not the same nominal
diameter or the same
nominal thickness as the pipe to be
examined.
In
addition, the calibration block did not contain calibration notchreflectors as
required. Each of the calibration block criteria described above are required
by ASME Section XI, Appendix III, supplement 7.
900202022) 9011-7
0
II
2
Unresolved
Item
50-269,270,287/89-37-01,
paragraph 2.a.,
documents
that
ultrasonic examinations performed on dissimilar metal welds at Oconee are
conducted using the shear wave
mode of sound transmission (shear
wave
transducers).
As
a result of the intergranular stress corrosion cracking
problem experienced on boiling water reactors, longitudinal wave transducers
have been found necessary to properly examine dissimilar metal welds if Inconel
base material is involved or if
alloy 82 or 182 are used as a wel.d butter or
filler material.
At the conclusion of the inspection, the licensee had not been able to identify
the materials in the dissimilar metal welds at Oconee.
The licensee also
.stated that, with the exception of the centrifugally cast stainless steel
reactor coolant piping at McGuire and Catawba,
all dissimilar welds at these
sites were also examined with shear wave transducers.
0Il
REPORT DETAILS
1.
Persons Contacted
Licensee Employees
- B. Foster, Maintenance Superintendent
- A. Gladney, Corporate Quality Assurance
- W. Hunt, Corporate Quality Assurance
- E. LeGette, Compliance Engineer
- E. Miller, Corporate Quality Assurance Technical Superintendent
- R. Morgan, Site Quality Assurance Manager
- R. Sweigart, Acting Station Manager
- T. Tucker, Corporate Quality Assurance
Other licensee employees contacted during this inspection included
craftsmen, engineers, technicians, and administrative personnel.
NRC Resident Inspector
- L. Wert, Oconee Resident Inspector
- Attended exit interview
2.
Inservice Inspector (ISI) Observation of Work and Work Activities -
Unit 3
(73753)
The inspector observed the ISI work and ISI-related activities, indicated
below, to determine whether examination activities performed on systems
and components containing reactor coolant, inside containment, were being
conducted in accordance with applicable procedures, regulatory require
ments,
and licensee commitments.
The applicable code for ISI is the
American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME
B&PV)
Code,
Section XI,
1980 edition with addenda through Winter 1980.
Duke Power Company (DPC)
nondestructive examination (NDE)
personnel were
performing the liquid penetrant (PT), magnetic particle (MT), radiography
(RT),
visual (VT),
and the ultrasonic (UT)
examinations with Babcock and
Wilcox (B&W)
personnel providing technical assistance to DPC personnel
during the UT of components not totally familiar to the DPC personnel.
Steam generator tubing eddy current (EC)
examination data collection was
being accomplished by DPC personnel with (B&W) and
personnel
performing independent analysi.s of the data.
The inspector observed the following three methods of examination to
determine if the ISI-related activities were conducted in accordance with
approved procedures by qualified and certified personnel knowledgeable of
the examination methods,
selection, and operation of the test equipment
for specific materials and components.
2
a.
Volumetric examination of welds using the manual (A-Scan) technique
The inspector observed calibration activities and portions of the
in-process ultrasonic examinations being conducted
on the items
indicated below.
Procedure
Calibration
Item and ID Number
Number
Block
Component
B05.050.012 (3PSP-1)
ISI-120
40367
Pressurizer Spray
Terminal
End
to
Safe-end
B05.050.012A (3PSP-1)
ISI-120
40406
Pressurizer Spray
Terminal
End to
Safe-end
B09.011.081 (3PSP.3)
ISI-120
40406
Pressurizer Spray
Piping
B05.050.011 (3PHB-17)
ISI-120
40414
Pressurizer Surge
Nozzle to Safe-end
B05.050.009 (3PSL-10)
ISI-120
40414
B-Hotleg Surge
Line Nozzle
B09.011.101 (3PSL-01)
ISI-120
40399
Pressurizer Surge
Terminal End
B06.040.001 (3RPV
Ligaments)
ISI-104
40390
Reactor Vessel
B02.040.004 (ISI
OCN3-004)
ISI-130
40305
Lowerhead to Tube
Sheet
B03.130.004 (3SGB
WB50-1)
ISI-130
40305
3B (Y-Z Axis)
Outlet Nozzle
B03.140.004 (3SGB
WB50-1)
ISI-130
40305
3B (Y-Z Axis)
Nozzle Inside
Radius
B03.140.003 (3SGB
WB50-2)
ISI-130
40305
3B (W-Z Axis)
Nozzle Inside
Radius
Procedure Calibration
Item and ID Number
Number
Block
Component
(cont'd)
B03.130.003 (3SGB
WG50-2)
ISI-130
40305
3B (W-Z Axis)
Outlet Nozzle
B05.050.011A (3PHP-17)
ISI-120
40399
Surge Nozzle
Safe-end
During the inspector's observation of calibration activities, two
areas of concern were identified. The first concern dealt with one
of the calibration blocks (Cal.
Blk
No.
40367)
that was to be
utilized for the examination of dissimilar metal weld 3PSP-1,
Item
No. B05.050.012.
The inspector noted that calibration block
number 40367 had apparent surface cracks and material separations on
its outside diameter surface.
In addition, the block was not the
same nominal diameter or the same nominal thickness as the calibra
tion block that was going to be used on the opposite side of weld
(3PSP-1)
Item No.
B05.050.012A or the piping to be examined.
These
concerns were discussed with the B&W examiner who was acting as a
technical adviser to Duke and who subsequently noted that the block
also did not have notches that were required by ASME Section XI,
Appendix III, Supplement 7, and the examination procedure.
The inspector reviewed the Oconee Unit 3, 2nd Interval,
Inservice
Inspection Plan and discovered that the calibration block had been
improperly designated
as the calibration block to
use for the
examination of Item
No.
805.050.012.
Duke
Power Company,
Quality
Assurance Program,
Procedure QA-513,
Rev. .6, states, in part, that
the Quality Assurance Engineer shall
prepare each
Inservice
Inspection Plan and that each plan shall include a detailed listing
of welds or components to be examined including their classification,
category and item number, required examination procedure, and
calibration standards to be used. In addition, paragraph 5.6 of the
Quality Assurance
Program
Procedure
requires that the Quality
Assurance Engineer or his designee,
in writing, shall review each
Inservice Inspection Plan or Summary Report for conformance to the
ASME Section XI, Final Safety Analysis Report, Technical Specifica
tion, and other licensing commitments.
The finding described above
was reported to Oconee's management as Violation 50-287/89-37-02,
"Inadequate
Technical
Review of Oconee's
2nd Interval
Inservice
Inspection Plan."
The
second area of concern observed,
during the calibration of
dissimilar metal welds listed above, was that DPC and B&W ultrasonic
examiners were using the shear wave mode of sound transmission (shear
wave transducers) to examine dissimilar metal welds at Oconee Nuclear
Station.
As a result of the intergranular stress corrosion cracking
4
problems experienced
on Boiling Water
Reactors in the early to
mid-1980's,
angle
beam
transducers which
produce
refracted
longitudinal wave mode of sound transmission have been found to be
necessary to properly examine dissimilar metal welds in some 600
series base materials or when alloy 82 or 182 are used as-a weld
butter,and/or filler material.
Upon finding that Duke
Power was using shear wave transducers to
examine all dissimilar metal welds at Oconee.
McGuire and Catawaba
Nuclear Stations, with the exception of the centrifugally cast
stainless steel reactor coolant piping to component nozzles at their
McGuire and Catawba Nuclear Stations, the inspector requested that
the licensee identify what materials were. involved in Oconee's
At the conclusion of the inspection, the
licensee had not been able to provide the inspector with this
information.
This item was
reported to Oconee's management
as
Unresolved Item 50-269,270,287/89-37-01,
"Ultrasonic Examination of
Dissimilar Metal Welds Using Shear Wave Mode Transducers."
b.
Radiocraphic .Review of 1ST Class 2 Piping Welds
Radiographs of the ISI welds listed below were reviewed by the
ins
or.
These radioQraphs were compared to the requirements of
the licensee's approved radiographic procedure, NDE-12, Rev. 7
Item No.
Weld No.
Size
C05.021.309
3-03-3FWD-74-A
24" dia. x 1.218" thk.
C05.021.360
3-01A-23-04
26" dia. x .875" thk.
C05.021.372
3-01A-24-02
26" dia. x .875" thk.
C05.021.378.
3-01A-24-03
26" dia. x .875" thk.
C05.022.013
3-01A-24-02L
26" dia. x .875" thk.
C05.022.009
3-01A-23-04L
26" dia. x .875" thk.
C05.022.015
3-01A-24-03L
26" dia. x .875" thk.
c.
Eddy Current Examination
The inspector
observed
performing
the
3600 rotating coil
examinations listed below.
With the exception of these enhanced
examinations, all
(SG)
tubes designated
for
examination with the bobbin coil (9210 in SG-A and 9226 in SG-B) had
been completed and all eddy current data evaluated by B&W and Duke
Power.
S G
Row No.
Column No.
Position
B-Hotleg
120
105
8th support plate
B-Hotleg
122
103
8th support plate
B-Hotleg
123
103
14th to 15th support
plate
B-Hotleg
82
124
14th support plate
The inspector reviewed qualification documentation for a select sample of
examiners audited during the examinations previously observed.
The
qualification and certification records for the following personnel were
reviewed:
Company
Examiner
Method-Level
H.A.D.
RT/Level II
G.E.H.
UT/Level II
J.S.S.
EC/Level IIA
J.L.R.
PT/Level II
Within the areas examined, violations or deviations were not identified
except as noted in paragraph 2.a. above.
3.
Review of Oconee Unit 3 Request for Relief No. 89-08
Duke
Power Request for Relief No.
89-08,
dated October 13,
1989,
was
reviewed by the inspector.
This untimely relief request addressed
reoortable indications found in reactor coolant pumps 3A2 and 3B1 main
flange stud hole threads. The damage to the stud hole threads was not due
to cracks in the base metal or pitting caused by boric acid corrosion but
instead was
more characteristics of damace dore durig the process .01
removing the studs.
The damage to the five threads on reactor coolant
pump casing 3A2 was typical of damage seen after removing a stuck stud.
The damage to the three threads on reactor coolant pump casing
3B1 was
tyical
of damace due to stud handling during removal and installation.
The discrepant conditions were found ouring tne 9th outage for Unit 3 in
the Spring of 1987. Oconee Unit 3 is presently in the 11th outage.
At
the time of discovery, engineering justifications were written and the
condition on both main flanges accepted based on the fact that a main
flange stud would be installed in each of the two holes in question and
tensioned normally.
Although credit would not be taken for these studs
for operability purposes,
the licensee felt that. their actions were
conservative based on a letter from B&W (Serial
No.
SGBM-85-507,
dated
July 9, 1985),
stating that damaged threads in one stud hole would not
affect the safety or operability of a reactor coolant pump since each pump
has been evaluated for 19 of 20 studs in-place.
However,
the inspector
was concerned over this relief request for the following reasons:
a.
Why was the condition not reported by a relief request in 1987 when
it was found?
b.
Why did drawings accompanying the Relief Request have handwritten
notes on
them indicating that Fermanite
may
be used to seal the
.The inspector held discussions with the Supervisor of Maintenance Engineer
ing, the Maintenance Engineer for the reactor coolant system, and the ISI
6
Coordinator
(QA
Engineer)
and reviewed
documents
provided by these
individuals. The licensee's responses to the above inspector's questions
are as follows:
a.
The licensee did not report the discrepant conditi.ons in 1987 because
maintenance engineering and the ISI QA Engineer at the time concluded
that,
since
the discrepancies were not
service-induced but
mechanically induced, they were not reportable to ASME Section XI.
b.
Discussions with the Supervisor of Maintenance Engineering revealed
that the handwritten notes on the licensee's drawings concerning
fermanite were no longer applicable.
During the 1982 to 1985 time
period, all Bingham reactor coolant pumps at Oconee had the pump
casing seat re-machined and a new stainless steel/graphite
typ'e
gasket was installed in lieu of the old stainless steel/asbestos type
gasket previously used. Fermanite injection ports were also removed
durino the above pump casing refurbishing efforts and set screws were
installed in their place.
In the time period since this work was
performed,
no leaks have been observed in the pump casing sealing
surface.
The
NRC Office of Nuclear Reactor Regulations (NRR)
will evaluate the
licensee s request for relief from the ASME
Code requirements.
However a
complete copy of the B&W evaluation analysis (Document
No.
31-1153263-00)
should be forwarded to NRR for their review as required by ASME Section XI
Paragraph I WB-3125.
Within the areas examined, violations and deviations were not identified.
4.
Exit Interview
The inspection scope and results were summarized on December 1, 1989, with
those persons indicated in paragraph 1.
The inspector described the areas
inspected and discussed in detail the inspection results listed below.
Proprietary information is not contained in this report.
During discussion of the violation listed below,
the Corporate Quality
Assurance Technical Superintendent requested that the Inspection Report
reflect that a Babcock and Wilcox examiner discovered concurrently with
the inspector the fact that the calibration block did not have calibration
notch reflectors.
(Open) Violation 50-287/89-37-02, Inadequate Review of Unit 2, 2nd Period,
2nd Interval, Inservice Inspection Plan, paragraph 2.a.
(Open) Unresolved Item 50-269,270,287/89-37-01, Ultrasonic Examination of
Dissimilar Metal Welds Using Shear Wave Mode Transducers, paragraph 2.a.