ML18037A428
| ML18037A428 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/09/1993 |
| From: | Blake J, Chou R, Coley J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18037A426 | List: |
| References | |
| 50-259-93-26, 50-260-93-26, 50-296-93-26, NUDOCS 9308250138 | |
| Download: ML18037A428 (18) | |
See also: IR 05000259/1993026
Text
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Report Nos.:
Licensee:
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIET1'ASTREET, N.W., SUITE 2900
ATLANTA,GEORGIA 303234199
50-259/93-26,
50-260/93-26,
and 296/93-26
Valley Authority
6N 38A Lookout Place
1101 Market Street
Chattanooga,
TN 37402-2801
Docket Nos.:
50-259,
50-260,
and 50-296
License Nos.:
and
Facility Name:
Browns Ferry 1, 2,
and
3
Inspection
Conducted:
July 12-16,
1993
Inspectors:
c,W
J.
L.
Co ey Jr.
Date Signed
R.
C.
Chou
7-M79 3
Date Signed
Approved by:
J. J.
ake,
Chief
at
ials
and Processes
Section
E
ineering Branch
Division of Reactor Safety
Date Signed
SUNMARY
Scope:
This routine announced
inspection
was conducted
in the areas of inservice
inspection (ISI) - observation of work and work activities
and review of
completed records
and evaluations,
review of radiographic film, verification
of pipe support modifications, review of pipe support calculations
and
verification of corrective actions taken
on previous enforcement
items.
Results:
ISI invessel
weld repair work on the core spray sparger
T-box piping and steam
dryer
and subsequent
visual inspection
were being conducted
by General
Electric
(GE) in a effective manner.
In process
underwater welding on the
steam dryer exhibited good physical characteristics
and met minimum
dimensional
requirements.
9308250138
9308l2
'PDR
ADOCK 05000259,'.(t
.6
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PDR-. ~r-'
Workmanship
on new support modifications were good
and met minimum dimensional
requirements.
.Design calculations for support modifications exhibited good
quality, with the exception of the improper handling of the revised stress
loads
by the pipe support design
group (Violation 50-259,260,296/93-26-02,
paragraph
5).
Radiographs
of welds in the reactor water cleanup
system for Unit-2 revealed
a
weakness
which was discussed
with senior
TVA management
in that, the licensee
has allowed radiographic techniques
and examination
parameters
in the field
which are producing very minimal radiographic sensitivity for small bore/thin
wall piping.
TVA's manager of inspection services
reviewed the applicable
film and agreed with the inspector's
assessment.
The licensee
also agreed to
improve the radiographic sensitivity by utilizing a combination of techniques
which include using higher
R factor film (slower film), longer-source-to-film
distances,
and/or Level III film interpreter rejection of minimal radiographic
quality techniques.
Corrective action for two previously open enforcement
items in the area of
welding were reviewed.
One violation which was addressed
by TVA's welding
engineering
group was handled appropriately.
The corrective action for the
other violation was handled
by a contractor
and was limited in scope,
not
presently
implemented,
and will not prevent reoccurrence
of the discrepancy
by
the contractor (Violation 50-259,260,
296/93-26-01,
paragraph T.B.[2]).
Preparations
for the September
1993, Unit 3, reactor vessel
examinations
from
inside the reactor vessel
are nearing completion.
Discussions with cognizant
TVA and
GE personnel
revealed that all examiner training has
been
completed
for these
examinations.
Applicable technical
procedures
are also presently
receiving their final review.
This attempt
by a
BWR owner to perform beltline
ultrasonic examinations
from inside the reactor
vessel,
as well as conducting
the associated
examiner training has
been handled extremely well by the
licensee
and
GE.
In the areas
inspected,
two violations were identified,
one for failure to
follow design
procedures
when performing design calculations
and the other for
failure to take adequate
corrective action
on a previously identified welding
program enforcement
item,
no deviations
were identified.
REPORT DETAILS
Persons
Contacted
Licensee
Employees
- H. Bajestani,
Technical
Support
Manager
- S. Bugg, Environmental
and Waste Control Manager
- C. Crane,
Maintenance
Manager
- H. Crisler,
Lead Engineer,
Site Engineering
- P. Dey, Civil Engineer
- K. Grooms, Supervisor,
Site Welding Engineering
- H. Herrell, Operations
Manager
- J. Johnson,
guality Assurance
Manager
- J. Haddox,
Manager,
Nuclear
Engineering
- P. Osborne,
Acting Lead Civil Engineer
- J. Rupert,
Engineering
and Modifications Manager
- J. Sabados,
Chemistry Manager
- P. Salas,
Licensing Manager
- H. Turnbow, Inspection Services
Manager
- R. Wells, Compliance Licensing Manager
- J. Whitaker, ISI Level III Examiner
- 0. Zeringue, Site Vice President
Other licensee
employees
contacted
during this inspection
included
craftsmen,
engineers,
technicians,
and administrative personnel.
Other Organizations
- G. Nelson, Project Manager,
GE Nuclear Services
NRC Resident
Inspectors
- T. Liu, Intern, Resident
Inspector
- J. Hunday,
Resident
Inspector
- C. Patterson,
Senior Resident
Inspector
- G. Schnebli,
Resident
Inspector
"Attended exit Interview
Inservice Inspection - Unit 3
Background
Browns Ferry Unit 3 is in an extended
shut
down status,
in cycle 5B, of
the third 40-month period, of the first ten-year
interval.
Unit 3
received its Operating
License
on August 18,
1977.
(While the ten-year
inspection interval would normally end
on the tenth anniversary of the
date of commercial
operations,
this unit is still considered
to be in
the first, ten-year ISI inspection interval to compensate
for this
extended
outage.)
The applicable
code for ISI, for Unit 3 is the
ASHE B8PV Code,
Section
XI, 1974 Edition with Addenda through the
Summer
1975
(74S75) for
everything except technique.
The applicable
code for technique is the
ASHE BKPV Code,
Section XI, 1986 Edition and
no Addenda.
For the currently scheduled
(September
16,
1993)
RPV inspections,
ASNE
BEPV Code Section XI, 1989 Edition with Addenda through
1991, will be
used.
This Code implements
Appendix VIII "Performance
Demonstration
For
Ultrasonic Examination Systems",
however, for the Unit 3 examinations,
Appendix VIII was
used for guidance only.
'a ~
Observation of ISI Work and Work Activities (73753)
The inspector
observed
GE performing in-process
underwater
repair activities,
and the subsequent
visual examinations
on the
internal reactor vessel
components listed below.
These work
activities were reviewed to determine whether the approved
applicable procedures
and technical
instructions
were being
followed, if welding exhibited good physical characteristics,
and
if visual examinations
revealed that minimum dimensional
requirements
were being met.
Welding and visual examinations for
the following welds were verified:
Meld Identification
Size
Com onent
Examined
b.
RWR-3-383592-G006
1/4"x 76"
Steam Dryer 8
230'WR-3-383592-G007
1/4"x 76"
Steam Dryer 8
310'he
above work was being performed in accordance
with DCN ¹
W179328.
The inspector also reviewed completed
DCN ¹ W18096A,
which welded bracket assemblies
on both of the core spray piping
T-box headers.
This repair modification was performed to resolve
concerns that
a previously identified crack indication will not
cause
complete separation of this piping.
Review of Completed
ISI Records
and Evaluations
(73755)
The inspector -reviewed the 1st- interval -inspection
program,
the
Unit 3 inspection
outage plan,
and system drawings, to select
a
representative
sample of examination reports for determining
whether the ISI files are complete
and the data is within the
previously established
acceptance criteria.
Examination data for
the following welds
was reviewed:
Wld ld" if'
~d
Si*
~0
RCRD-3-OOOOC4
RCRD-3-000044
RCRD-3-000052
DCS-3-000014
DSC-3-0000410
THPCI-3-00062
RECIRC-GR-00000027
RECIRC-GR-00000028
DRHR-3-000005
DRHR-3-0000013
DSRHR-3-00005
C-F
B-J
B-F
B-J
B-F
B-J
B-J
B-J
B-J
B-J
B-J
6"
Dia. ISI-0043
4"
Dia. ISI-0332
4"
Dia. ISI-0332
12"
Dia. ISI-0331
12"
Dia. ISI-0331
14"
Dia. ISI-0333
28"
Dia. ISI-0328
28"
Dia. ISI-0328
24"
Dia. ISI-0330
24"
Dia. ISI-0330
24"
Dia. ISI-0330
The above records
were also reviewed to determine the following:
(1)
Whether the method, extent,
and technique of examination
comply with the licensee's
ISI program
and applicable
procedures.
(2)
If the examination data
was within the acceptance
criteria,
and whether the recording, evaluating,
and dispositioning of
findings are in compliance with the applicable
procedure.
(3)
Whether
NDE examiners
were adequately qualified to perform
their assigned
task.
(4)
Whether the method
used for NDE was sufficient to determine
the full extent of indication or acceptance.
(5)
Whether the licensee is utilizing the services of a third
party inspection
agency,
as required.
Based verification of the above weld examination
documentation
and
personnel certification records,
the inspector
concluded that ISI
records
are being completed
and controlled properly by TVA.
Review of Preparations
for the Unit 3 Reactor
Pressure
Vessel
Examinations
Conducted
In Accordance
With NRC's
New Rule For
Augmented Examination Of The Reactor
Pressure
Vessel
f10 CFR
50 55a (g)(6)(ii)(A)I
During-this inspection the inspector held discussions
with TVA and
GE to determine
when the Unit 3 reactor pressure
vessel
examinations
were presently
scheduled to be performed
and to find
out the status of prerequisite
issues
and processes
discussed
in
NRC, Region II Reports
Nos. 259,260,296/92-40
and 92-42.
These
prerequisite's
included successful
completion of examiner
performance
demonstrations
test for ultrasonic detection
and
sizing of indications,
equipment characterization,
approval of an
augment calibration response verification method
by the Authorized
Nuclear Inspector, drilling of new near surface
holes in the
reactor vessel
calibration block,
and development of
nondestructive
examination
procedures
for detection
and sizing of
indications.
As a result of the above discussions
the inspector discovered that
the reactor pressure
vessel
examinations
are presently
scheduled
for September
16,
1993.
The status of training, equipment
characterization,
calibration block modification,
and procedure
development
is on schedule.
The licensee
and
GE have handled the
prerequisite for this new examination effort extremely well.
Within the areas
examined,
no violation or deviation was identified
Review of Radiographic
Film - Unit 2 (57090)
The inspector reviewed
a sample of radiographic film and associated
records to determine whether they were prepared,
properly evaluated,
and
maintained in accordance
with the TVA's approved radiographic
procedure
N-RT-1, revision
16 and the applicable
ASNE Code.
Each radiograph
in
the weld film packages listed below was specifically reviewed to
determine whether the following examination
parameters
had been
correctly adhered to; penetrameter
type, size,
placement,
and
sensitivity; film density
and density variation; film identification;
film quality; and weld coverage.
Radiographs for the following welds
were .reviewed:
Weld Identification
Size
Comments
RWCU-2-004-G024
RWCU-2-004-G043
RWCU-2-004-G044
RWCU-2-004-G045
RWCU-2-004-G046
RWCU-2-004-G051
RWCU-2-004-G063
RWCU-2-004-G064
RWCU-2-004-G065
RWCU-2-004-G066
RWCU-2-004-G068
RWCU-2-004-G069
6"Dia.X
3"Dia.X
3"Dia.X
3"Dia.X
4"Dia.X
6"Dia.X
4"Dia.X
4"Dia.X
4"Dia.X
4"Dia.X
4"Dia.X
4"Dia.X
.432"
.300"
.300"
.300"
.337"
.432"
.337"
.337"
337II
.337"
33711
337
4T Holes Very Faint
4T Holes Very Faint
ID Dry Label
Added
ID Dry Label
Added
ID Indelible Ink
4T Holes Very Faint
Sat.
ID Dry Label
Added
ID Dry Label
Added
Sat.
Sat.
Sat.
Sat.'uring
the above verification reviews the inspector
observed that,
radiographic techniques
used in making the film exposures
were producing
radiogr aphs with very minimal radiographic sensitivity.
TVA's manager
of technical
services
subsequently
reviewed the applicable radiographs
and agreed with the inspector's
assessment.
As result, the manager of
technical
services
also agreed to improve the radiographic sensitivity
by utilizing a combination of techniques
which include using higher
R
factor film, longer-source-to-film distances,
and/or Level III film
interpreter rejection of minimal radiographic results.
These corrective
measures will be verified during
a subsequent
inspection.
Within the areas
examined,
no violation or deviation
was identified.
Pipe Support
Walkdown Reinspection
- Unit 3
The inspector
randomly selected
13 pipe supports
which were modified
between
August
1992
and present
time and
had previously been
accepted
by
the licensee.
The
13 pipe supports
were all in large bore piping for
three different safety-related
systems
located outside of the drywell.
The walkdown reinspection
was completed with assistance
from the
licensee's
engineers
and
gC mechanical
inspector
who was also qualified
as
a welding inspector.
The supports
were partially reinspected
against
detail drawings,
including the original walkdown sketches,
the design
change notices
(DCN's), the field design
change notices
(FDCNs),
and the
final as-built drawings (for some supports).
They were checked for
configuration, identification, fastener/anchor
installation,
anchor
size,
anchor type,
anchor marking,
anchor
edge distance,
base plate size
and thickness,
plate warpage,
member size,
weld sizes,
component
identification numbers,
component sizes
and settings,
dimensions,
oxidation accumulation,
maintenance,
and damage protection.
The
supports
reinspected
during the current inspection
are listed below.
All the support modifications reinspected
were found to be acceptable.
Table 4-1
Walkdown Reins ection
Su
orts
Item No.
Su
ort No.
Revision
No.
1
2
3
4
5
6
7
8
9
10ll
12
13
3-17B300-280
3-17B300-285
3-17B300-293
3-17B300-294
3-17B300-299
3-47B450-315
3-47B450-316
3-47B451-R0058
3-47B451-R0066
3-47B451-R0094
3-47B451-R0095
3-47B451-R0096
3-47B451-R0098
DCA-001
DCA-001
DCA-001
DCA-001
DCA-001
DCA-001
CCD-000
DCA-000
DCA-001
DCA-000
DCA-000
DCA-000
DCA-001
Within the areas
examined,
no violation or deviation
was identified.
Pipe Support Calculation
Review - Units 2"and
3
The design calculations listed below in Table 5-1 were partially
reviewed
and evaluated for thoroughness,
clarity, consistency,
and
accuracy.
The following design criteria and standards
were used to
compare the design calculations:
~Cate or
Criteria
Standard
Document
No.
BFN-50-C-7107
DS-C1.7. 1
Rev.
No.
Title
6
Design of Class I
Seismic
Pipe
and
Tubing Supports
General
Anchorage to
Concrete
Standard
DS-C1.7.3
Concrete
Anchorage
Application of
Baseplate II
The above calculations
contained the purpose,
assumptions,
references,
computer programs,
design input, graphics,
main body calculations,
summary of results,
conclusions,
and attachments.
The computer programs
used included:
FAPPS program for the structural
member analysis;
Baseplate II program for the flexible base plate
and
anchor bolt analysis;
DD Lug design
program for the integrated
lug
design
and analysis;
and
Conan program for the G-32 anchor bolt spacing
violation analysis.
The attachments
included existing pipe support configuration from
walkdowns,
proposed
support modifications or design
change notices
(DCNs),
Employee concerns checklist,
and computer input and output for
frame and base plate analysis.
The review included:
Design loads from
the latest revised stress
calculations,
overall calculation contents,
allowable loads
used,
over stress
condition justification if any,
standard
component capacity
and settings,
weld sizes
and symbols,
deflection, bolt sizes
and length
, base plate flexibility, and bolt
spacing violations.
Table 5-1
Su
ort Calculations
Reviewed
Su
ort No.
Unit
Calculation
No.
Rev.
Comments
2-47B452S0245
2-47B452S0246
2-47B452R0056
2-47B452S0247
2-47B452S0248
N1-370-56R-0004
J
2
2
2
2
2
3
CD-02074-894001
CD-02074-894002
CD-Q2074-894003
CD-Q2074-894004
CD-Q2074-894005
CD-Q3070-922408
2
2
2
2
- See
Note
2
- See
Note
I
'
N1-370-56R-0005
3
CD-03070-922408
1
.N1-370-57R-0002
.3
CD-.03070-922410
3
N1-370-57R-0003
3
CD-03070-922410
3
N1-370-57R-0004
3
CD-03070-922410
3
- Note; The support calculation
was not updated
and revised to reflect
the latest revised stress
loads.
In general,
the design calculations
were of good quality except
as noted
in Table 5-1 above.
Supports
2-47B452S0247
and 2-47B452S0248 of
calculations
CD-92074-894004
and -894005 were two snubbers
for two
direction restraints
acting at the
same location for the node point R57
as specified in stress
calculation
CD-92074-893811.
Both of the support
calculations
and the stress
calculation were generated
in August,
1989.
The stress
calculation
was revised to Rev.
3 on Hay 15,
1990 and the
stress
loads at Node
R57 at X-direction were revised to increase
about
22% from the faulted loads of 12720 lbs to 15616 lbs.
The corresponding
pipe support calculation CD-f2074-894005
was not updated
and revised to
qualify the
new load increase.
This support calculation
was revised to
Rev.
2 on July 12,
1993, during the beginning of this inspection.
The
stress
load for the other support
(2-47B452S0247)
was not revised in the
stress
calculation.
After the inspector revealed this problem to the licensee,
the
licensee's
engineers
took quick action to revise the stress
and two
support calculations to show both supports to be acceptable for the new
loads
based
on the
DCN.
The licensee
revised the stress
calculation
and
removed the increase
loads
back to the original loads
shown
on the
stress
calculation
based
on the statement that the resultant
loads will
not change
due to the snubber rotations.
The loads in the individual
snubber will be changed
and qualified.
The licensee
also reviewed all
support calculations
contained in this stress
calculation to see if any
other support calculation
had the
same
problem of not updating the
support calculations for the latest revised
new stess
loads.
A Problem
Evaluation Report
(PER)
No.
BFPER930088
was issued
by the licensee for
investigating the root cause.
10 CFR 50, Appendix B, Criteria III,
Design Control requires that,
design
changes
shall
be subject to design
control measures
commensurate
with those applied to the orginal design.
TVA Nuclear
Engineering
Procedure
NEP-3. 1, Attachment 4,
Page
1 of
1
states that design input, including information such
as loads,
temperature,
etc. shall
be current,
referenced,
and applied.
Rigorous Analysis Checklist requires that the correct support loads from
the post processor
output, or adjusted
loads from hand calculations,
have
been transmitted to the support designer.
The two support
calculations
CD-(2074-894004
and -894005 were not updated
and revised to
reflect the latest stress
loads or the
new loads
due to the snubber
direction changes.
This item is identified as Violation 50-
259,260,296/93-26-02,
"Failure of Updating the Pipe Support Calculations
for the
New Stress
Loads.
In addition, this violation is
a repeated
case.
A similar discrepancy
case of not updating the pipe support calculation per the latest stress
calculation
was documented
in Inspection
Report
No. 50-259,260,296/92-38
for Unit 2 calculations
and
was cited as
a violation.
The licensee
claimed in their response
to the previous violation that it was
an
isolated
case
but the
same
problem keeps reoccurring.
The licensee
needs to review the problem and find a solution that will prevent its
reoccurrence.
The snubber for Support
No. 2-47B452S0245
and calculation
No. CD-g2074-
894001
Provided the restraint in Global
X -direction which was East-
West direction
as specified in the stress
Calculation
No. CD-g2074-
893811,
Rev. 3.
The allowable loads of Snubber
HSSA-10, Stroke 6" from
Bergen Patterson
Catalog
66R is 10,000 lbs for upset condition
and is
greater than the applied load of 8328 lbs which was stated at faulted
condition per stress
calculation
No. CD-(2074-893811,
Rev. 3.
Therefore,
the snubber
and support is acceptable.
Support
No. 2-
47B452S0246
(calculation CD-(2074-894002,
Rev.
2) used
a
1 1/4" thick
base plate
and qualified the base plate
on sheet
9 by using the formula
specified in TVA Civil Design Standard
DS-C1.7.1,
Section 5.1.1.
The
formula uses
anchor bolts
as tension axis
and the intersection
(the
first contact) of base plate
and concrete
as compression
axis.
The bending
arm distance
between
compression
axis
and tension axis is
defined
as
2 times baseplate
thickness
plus the distance
from the
compression
edge of attachment to the tension bolts.
This formula or
the bending
arm distance
was established
and verified by the supporting
calculation
CSG-85-002,
Rev. 0, dated July 31,
1985.
This supporting
calculation justified the formula's accuracy
and conservatism for the
anchor bolt design,
by comparing the results of this formula and the
output of the
CDC Base Plate II Finite Element Analysis.
This formula
accounts for the prying action or the base plate flexibilityanalysis.
It does not use the edge of base plate or the co'mpression
anchor bolts
as the compression
axis
as the normal rigid analysis for the base plate
or anchor bolt design.
Per
IE Bulletin 79-02, the licensee
was requested
to respond to the
bulletin by stating the base plate flexibilityconsiderations
in the
analysis
and calculations.
The inspector requested
the licensee to
provide the IE Bulletin response
and the example
1 and
3 of the
supporting calculation for review.
However, the information requested
could not be retrieved in the limited time available.
Therefore, this
item is identified as Inspector
Followup Item (IFI) 50-259,260,296/93-
26-03,
"Review of the supporting calculation
and
IE Bulletin Response
for the Base Plate Flexibility".
Calculation CD-(3074-922408,
Rev. 1 qualified the pipe supports
Nl-370-
and -0005 by using the deflection criteria 1/8".
The maximum
deflection was 0.071" which was less
than the 1/8" allowable deflection
specified in Section 1.4.2.13.b of Browns Ferry Nuclear Plant Design of
Class
.1 Seismic
Pipe
and Tubing Supports,
Rev. 6.
4'
These
two supports
are small bore supports
and tied back to the pump.
They were not the restraints
adjacent to
a
pump to restrain the
pump.
Therefore,
a deflection of 1/16" limit based
on Section 1.4.2. 13.a was
not applied.
With the areas
examined,
one violation was indentified,
no deviation
was
identified.
6.
Review of Work Plan - Unit 2
During the review of the pipe support calculations,
some of the
calculations
required the supports to be modified before restart.
The
inspectors
randomly selected
four pipe supports
(2-47852S245,
2-
478452S0247,
2-478452S0248,
and 2-478452R0056)
and reviewed their work
plan to .see if-the modifications for those supports
had been
completed
by the construction
group
and accepted
by the
gC inspectors.
After
reviewing the work plan, the inspector determined that all the supports
were adequate.
They were modified by the construction
group and
accepted
by gC inspectors.
Within the areas
examined,
no violations or deviations
were identified.
Licensee Corrective Action on Previous
Inspection Findings
(92702)
(Closed) Severity Level
IV Violation 50-259,260,296/93-05-01,
"Failure of Welder to Follow the Parameters
of His gualified
Welding Procedure Specification"
Corrective measures
addressed
in the licensee's letter of response
dated April 21,
1993,
were verified by the inspector.
These
corrective measures
were complete,
very effective,
and should
prevent the re-occurrence
of this discrepancy.
B.
(Open) Severity Level
IV Violation 50-259,260,296/92-43-01,
"Failure to Follow Installation Specification Requirements"
The corrective measures
committed to by the licensee
in their
letter of response
dated
February
10,
1993, consisted of:
(1)
Having
GE assemble
a documentation
package containing the
expertise
and experience of personnel
performing weld repair
overlays
and submitting this information to TVA's Nuclear
Engineering
Department for review in lieu of performing weld
overlay mockup training.
The inspection
noted that
GE had
submitted this package to TVA, aqd
TVA had approved the
package.
However,
when the inspector requested
the package
in order to verify the welders expertise
and experience
he
was informed that the documentation
was in storage off site.
Therefore, this item will remain
open until the package
can
be retrieved
and reviewed.
10
(2)
In addition to the above action taken to satisfy the welding
program nonconformance
to prevent reoccurrence
of a similar
discrepancy
GE initiated the development
and use of a
DCN
requirements
checklist.
This checklist identified the
requirements
in the
DCN and provided
a tracking method of
accounting for the completion of each requirement.
The
completed checklist
was then reviewed
and approved
by the
Project Manager
and GE's
gC group.
The effective date for
implementing the
DCN checklist form was January
5,
1993.
The inspector investigation into GE's implementation of the long
term corrective action revealed that, only one group within GE's
welding organization
implemented the
DCN requirement checklist.
This group subsequently
left the Brown's Ferry facility and the
checklist which was designed to prevent reoccurrence
of the
reported discrepancy,
presently is not used
by any other group
within GE's welding organization.
The inspector also noted that
GE had not revised their Welding Manual or any other technical
or
administrative procedure
to'implement the corrective action
commitment requiring the
DCN requirement checklist.
Inadequate
-corrective action to prevent the reoccurrence
of a significant
noncomformance
is a violation of 10CFR50,
Appendix B, Criterion
XVI, and was reported to the licensee
as Violation 50-
259,260,296/93-26-01,
"Failure of Licensee to Take Adequate
Corrective Action for a Previously Identified Vendor Melding
Program Nonconformance".
Within the areas
examined,
no violation or deviation was identified
except
as noted in paragraph
7.B.(2) above.
Exit Interview
The inspection
scope
and results
were summarized
on July 16,
1993, with
those
persons
indicated in paragraph
1.
The inspectors
described
the
areas
inspected
and discussed
in detail the inspection results listed
below.
Proprietary information is not contained in this report.
Dissenting
comments
were not received
from the licensee.
(Open) Violation 50-259,260,296/93-26-01,
"Failure of Licensee to Take
Adequate Corrective Action for a Previously Identified Vendor Welding
Program Nonconformance"
(Open) Violation 50-259,260,296/93-26-02,
"Failure of Licensee to Update
Pipe Support Calculations for New Stress
Loads"
(Open) Inspector
Followup Item 50-259,260,296/93-26-03,
"Review of
Supporting Calculation
and IE Bulletin Response for Base Plate
Flexibility"
0',