ML18036B079
| ML18036B079 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/18/1992 |
| From: | Blake J, Chou R, Coley J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18036B077 | List: |
| References | |
| 50-259-92-38, 50-260-92-38, 50-296-92-38, NUDOCS 9212020102 | |
| Download: ML18036B079 (17) | |
See also: IR 05000259/1992038
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323
Report Nos.:
50-259/92-38,
50-260/92-38,
and 50-296/92-38
Licensee:
Valley Authority
3B 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:
November 2-6,
1992
Inspectors:
.. Coley Jr.
.
C.
Chou
Approved By:
J
J
Blake, Chief
at rials and Processes
Section
Engineering
Branch
Division of Reactor Safety
Date Signed
II /& fc
Date Signed
lr (
Date Signed
SUMMARY
Scope:
This routine,
announced
inspection
was conducted
on site in the areas
of
inservice inspection (ISI) - observation
work and work activities; review of
radiographic film for class
1 reactor water clean
up
of Unit 1, shroud
manway access
hole cover, ultrasonic
(UT) examinations;
and
review of UT data (Information Notice [IN] 88-03
and
IN 92-57 "Cracks in
Shroud Support Access
Hole Cover Welds"); review of TVA actions with regards
to IN 92-35 "Higher Than Predicted
Erosion/Corrosion
in Unisolable Reactor
Coolant Piping Inside Containment";
review of TVAs purposed
response
to
NRC Generic Letter (GL) 88-01 Supplement
1
"NRC Position
on Intergranular Stress
Corrosion Cracking
(IGSCC) In
BWR Austenitic Stainless
Steel
Piping"; review
of previously open
NRC items;
and review of pipe support design calculations
for Unit 2.
9212020102
921123
ADOCK 05000259
8
Results:
One cited Violation No. 50-260,296/92-38-01,"
Inadequate
Design Controls for
Pipe Support Calculations,"
(paragraph
7);
one non-cited violation No.50-
260/92-38-02,
"Failure to Properly Identify Support Spring
Can Variability,"
(paragraph
7);
and
one unresolved
item No. 50-296/92-38-03,
"Evaluation of
Weld Conditions,"
(paragraph
3) were identified by the inspectors.
No
deviations
were identified.
One weakness
was also identified in the license's
evaluation of information Notice No. 92-35.
(paragraph
5)
Not withstanding
the items identified, the inspectors
concluded that
TVA management
is actively
involved in attempting to correct the root causes
of problems with hardware
and personnel
performance.
A balance of improved supervision,
personnel
training,
employee
commitment,
and improved procedures
appear to be obtaining
positive results.
0
REPORT
DETAILS
Persons
Contacted
Licensee
Employees
R. Baird, Principle Civil Engineer
- 0. Butler, Level III Examiner,
Inspection Services
- R. Cutsinger,
Lead Civil Engineer
- J. Davenport,
Regulatory Engineer
- S. Fox,
Level III Examiner,
Inspection Services
- F. Froscello,
ISI Supervisor
- E. Hartwig, Project
Manager
T. Knuettel, Licensing Engineer
- L. Madison, Supervisor,
Civil Engineering
- D. Massey,
Regulatory
Engineer
J.
McCord, Stress
Analyst
R. Phillips, Supervisor,
Material Engineering
- D. Nye, Recovery
Manager
G. Strickland, Material Engineering,
Corporate Office
- J. Sabados,
Chemistry
and Environment Manager
- M. Turnbow, Manager,
Inspection Services
- 0. Zeringue, Site Vice President
Other license
employees
contacted
during this inspection
included
engineers,
mechanics,
technicians,
and administrative personnel.
Other Organizations
General
Electric Nuclear Energy
(GENE)
T. Brinkman, Project Manager,
NDE Application Technology
H. Hart, guality Assurance
Manager
- R. Seals,
ISI Supervisor,
S. Stanford,
ISI Level II Examiner
NRC Resident
Inspectors
- J. Hunday,
Resident
Inspector
- C. Patterson,
Senior Resident
Inspector
- Attended exit interview
and initialisms used throughout this report are listed in the
last paragraph.
Inservice Inspection
- Observation of Work and Work Activities
Unit 3
(73753)
The licensee
is currently in the process
of preparing Unit 3 for
restart.
As part of this effort the licensee
has instituted
an
integrated
program to implement the commitments
made in response
to
NRC's position
on
IGSCC in boiling water reactors
(BWRs)
austenitic piping.
GL 88-01 requires mitigation of IGSCC in susceptible
piping by inspection,
repair
and or replacement.
To comply with the
requirements
of GL-88-01, the piping runs exposed to fluid temperatures
greater than
200 'F have
been replaced with type 316
NG (Nuclear Grade)
stainless
steel,
which is not susceptible
to
Piping runs not
exposed to fluid temperatures
greater
than
200 'F are being replaced
with type 316 stainless
steel.
To date,
the
12" dia.
and 20" dia. portions of the reactor recirculation
and the 6" dia. portions of the reactor water clean-up
(RWCU) systems
have
been replaced
inside containment.
The 28" dia. recirculation
piping that previously
had reported
IGSCC are in the process
of having
full structural
design weld overlays applied.
Installation
and welding of replacement
piping was accomplished
by
General
Electric
(GE)
Company,
Nuclear Services
and Project Department
under the umbrella of the
GE guality Assurance
(gA) Program.
After each
pipe weld is acceptable
by radiography,
mechanical
stress
improvement is
performed.
Nondestructive
examinations
(NDE), including preservice
examinations,
of the pipe welds are being performed
by TVA.
The applicable
Codes for the pipe replacement
project are:
American Society of Mechanical
Engineers Boiler and Pressure
Vessel
(ASME,
BEPV)
Code Sections III, V, and XI 1986 Edition.
ASME,
B&PV Code Section II and IX, latest edition.
The inspectors
observed
the preservice ultrasonic examination,
performed
on Weld No.
RWCU-3-001-G023, to determine
whether the approved
procedure
(N-UT-18, Revision
13) was being followed; whether the
examination
personnel
were knowledgeable of the examination
method;
and
whether the examiner
made the appropriate
interpretation of the test
results.
Within the areas
examined,
no violations or deviations
were identified.
Review of Radiographic
Film
Unit 3 (57090)
The inspectors
reviewed radiographic film, and associated
records,
for
Class I, Reactor
Water Clean-up
whether the
radiographs
were prepared,
evaluated,
and maintained in accordance
with
the applicable
Codes
(ASME Sections III, and V, 1986 Edition).
Included in this review was verification that the penetrameters
were the
correct type, size,
and properly placed
on the pipe,
and if adequate
sensitivity was obtained with the radiographic technique.
The
radiographs
were also reviewed to ensure that film density
was within
the allowable code variation, weld coverage
was complete,
and welding
discontinuities
were properly evaluated.
Radiographs for the following
welds were reviewed:
Meld Identification No.
RWCU-3-001-G001
RWCU-3-001-G011
RWCU-3-001-G014
RWCU-3-001-G015
RWCU-3-001-G016
RWCU-3-001-G017
RWCU-3-001- G018
RWCU-3-001-G019
RWCU-3-001-G020
RWCU-3-001-G021
RWCU-3-001-G022
RWCU-3-001-G023
RMCU-3-001-G024
RWCU-3-001-G025
RWCU-3-001-G026
Work Plan
No.
3250-92
3250-92
3539-92
3539-92
3539-92
3539-92
3539-92
3539-92
3539-92
3539-92
3539-92
3539-92
3634-92
3634-92
3634-92
Size
6"Dia.
6"Dia.
6"Dia.
6"Dia.
6"Dia.
6"Dia.
6"Dia.
6"Dia.
6"Dia.
6"Dia.
6"Dia.
6"Dia.
6"Dia.
6"Dia.
6"Dia.
The inspector's
review identified two weld radiographs
which
contained indications of welding conditions that the license
should
have investigated further.
The first weld was
RMCU-3-001-G024 which had three film segments,
(0-1,
2-3,and 3-0) where the consumable
"K" insert ring had not been
consumed.
In each
case the insert
had melted to the point that there
was
no lack
of fusion at either side of the weld prep but had not melted to the
point that the constituents
in the ring had flowed, (i.e. the insert
ring still maintained its original shape.)
The
Code does not provide
acceptance
criteria for incomplete insert melt,
so one option would be
to apply the criteria for elongated
indications.
If this criteria was
applied the weld would be rejectable;
however,
there
appears
to be
a
valid argument for not using this criteria since the ring did melt to
the point of fusion and therefore the weld joint may have the structural
soundness
required.
Cognizant
TVA management
was notified and although the licensee felt the
weld was acceptable
they offered to have
a metallurgical evaluation
made
of the weld soundness
with this condition.
In addition,
TVA offered to
research
ASHE Code
Cases
and
Code Inquiries,
as well as other industry
guidance,
in order to establish
procedural
guidelines which would give
definitive acceptance
criteria for this condition.
TVA actions with
regard to this weld are very responsible
since this is
a 316NG stainless
steel
weld which has received
mechanical
stress
improvement.
Any repair
to this material
would have
some detrimental effects
on its ability to
mitigate
The second
weld condition questioned
by the inspectors
was
on Weld No.
RWCU-3-003-G026.
The Radiographic
Inspection
Report
had
a comment adjacent to the evaluation for film segment
1-2 that
a
weld condition noted
on the radiograph
was
a weld cap condition.
Due to
the magnitude
(apparent
depth) of the indication, the inspectors
questioned
whether there might also
be
a lack of fusion indication,
on
the side of the weld prep edge,
along with the weld cap
edge condition.
During the discussions
with the licensee,
the inspectors
also learned
that the examiners
who had originally evaluated
the radiograph
had not
visually examined the weld to confirm the weld cap condition.
A visual
examination
was subsequently
performed
by the radiographic
examiner
and
the inspectors;
however,
the weld reinforcement
had
been
ground off for
ISI, which would have
removed
any weld cap indication.
Therefore,
cognizant
management
personnel
stated that another radiographic
shot
would be made of this film segment
in order to assure
that the
questioned
condition was not partly the result of lack of fusion.
The actions
proposed
by the license for the further evaluation of weld
conditions noted
by the inspectors
should
be adequate
to resolve whether
these conditions are acceptable
or rejectable.
This issue
however, will
be identified as unresolved
item 50-296/92-38-03,
"Evaluation of Welding
Conditions",
pending the results of the license's
evaluations.
Within the areas
examined,
no violation or deviation
was identified.
Observation of the Shroud
Nanway Access
Hole Cover Examinations
and
Review of Ultrasonic Data, Unit-1
(92701)
NRC,
IN 88-03 "Cracks in Shroud Support Access
Hole Cover Welds",
and
IN 92-57 "Radial Cracking of Shroud Support Access
Hole Cover Welds"
alerted licensees
of boiling water reactors
(BWRs) of the potential for
cracks in the welds of covers to the shroud support
access
holes within
the reactor vessel.
Each
BWR has
two access
hole covers in the shroud
plate,
one at
0 degrees
and the other at
180 degrees.
The access
hole
covers for the
Browns Ferry Unit
1 reactor were inspected
by UT
examination in late April, 1992.
Circumferential cracking
was detected,
and
a visual inspection
confirmed that the cracks
were through-wall.
At
that time the General Electric's
(GE's)
UT inspection fixture was not
configured to detect radial cracks,
but radial indications were detected
in the 0'ccess
hole cover during visual examination.
As a result
recommended
that additional
UT inspections
be performed from the inside
of the reactor with a modified
UT fixture for radial scanning.
In
addition
GE established
methods to examine the reactor vessel
and the
shroud
attachment
welds inside the vessel
from outside of the reactor
vessel,
in order to determine
the extent of the radial cracking.
On November 2,
1992, the inspectors
arrived at the Browns Ferry Nuclear
Plant to observed
the
UT examinations of the shroud
manway access
cover
plate welds
and to review the recorded data.
At this time
GE had
completed all scans
from inside the vessel
on the 0'anway
access
cover
plate weld, but had not fully reviewed or plotted the data.
Since
was in the process
of setting
up the Smart
2000 to perform the
examinations
on the 180'anway
cover from inside the vessel
the
inspectors
started their review with the recorded 0'ata.
This review
revealed that the scans
from the ledge side
and the cover side of the
manway access
cover had detected significant circumferential reflectors
l
i
indicative of IGSCC.
However, the data from the scans to detect the
radial indications
was inconclusive,
in that no apparent reflector was
producing
a signal indicative of IGSCC.
On November 3,
1992,
the inspectors
observed
the
UT examinations
performed
on the 180'anway
access
cover.
During these
examinations
the inspector
noted that the fixture used to detect radial indications
was limited in its ability to scan the entire circumference of the weld.
This was due to the limited room between
the access
cover weld and the
shroud
and reactor vessel
walls.
Both areas
prevented
the fixture from
scanning
the most susceptible
areas
to radial cracking.
In addition the
inspectors
noted from the
TV monitor which had
a camera
on the
fixture, that there
are almost
no parallel
surfaces
in the areas
adjacent to the wall of the vessel
or the shroud.
This ledge surface
condition would redirect the sound making the detection of radial cracks
extremely difficult from the inside of the reactor vessel.
The inspectors
noted
however that the visual radial indication
identified as
No.5 was entirely on the 0'hroud
access
manway cover.
The cover has parallel
surfaces
and there
are
no scan limitations.
Therefore, this indication should
be readily detectable
with UT.
However, the indication could not be detected
with the transducer
fixture configuration
used
by
GE while the inspectors
were at the site.
Review of portions of the 180'ccess
hole cover
UT data revealed
similar information to the 0'ata
in that, the
UT system easily
detected
circumferential indications indicative of IGSCC, but did not
detect
any apparent radial indications.
At the conclusion of the inspectors visit,
GE was still adjusting the
angles of the transducers
in the fixture in an attempt to detect the
visual radial indications.
The examinations
from outside the reactor
vessel
were not scheduled until the following week.
In addition to observing the ultrasonic examinations
on the 180'anway
cover
and reviewing portions of the data for both the 0'nd the
180'anway
covers,
the inspectors
reviewed examiner,
equipment,
and material
qualification and certification records,
and reviewed the following
procedures
to determined if their technical
content
was adequate
and
whether they had
been properly approved:
Procedure
ID & Rev.
GE-ADM-1002,
GE-ADM-1001,
Title of Procedure
Procedure for Review Process
and
Analysis of Recorded
Indications
Procedure for Performing Linearity
Checks
on Ultrasonic Instruments
GE-RDE-14-0488
0
Procedure for Inspection
and
Installation of Access
Hole Cover
Scanner
GE-UT-211
Procedure for Automatic Ultrasonic
Examination of the Shroud Support
Access
Cover Plate
GE-VT-202
Invessel
Visual Inspection
Within the areas
examined,
no violations or deviations
were identified.
Review of TVAs Actions Regarding
" Higher Than Predicted
Erosion / Corrosion in Unisolable Reactor Coolant Pressure
Bounty Piping
Inside Containment at
a
BWR" (92701)
The inspectors
reviewed licensee activities performed to date,
and those
planned for the near future, with regards to IN 92-35.
This review
revealed that
TVA intended to use the Electric Power Research
Institute
(EPRI)
Checmate
Computer
Program to select priority ranking for
components that would be examined
on the feedwater
system inside
containment
in response
to IN 92-35.
To date,
Pass
1 of Checmate
has
been performed
on the Unit 2 piping to
rank each
component for examination;
however,
TVA's site copy of IN 92-
35 was missing Attachment
1 to the IN, which is
a detail drawing of the
piping configuration
and the area
where the erosion/corrosion
was
occurring at the site which identified the problem.
A review of
checmate,
pass
1, ranking of the Unit 2 pipe component in question
revealed that this component
had not been
ranked
high enough to ensure
its examination
under TVA's present
planning.
In addition,
a review of
TVA's piping drawings revealed that, the Browns Ferry reactors
also
had
the
same piping configuration
as the plant with the reported condition.
The inspectors
discussed
the inspection findings with
cognizant
management
and engineering
personnel
and were assured
that the area of
concern for IN 92-35 will be examined next refueling outage for Unit 2,
and before startup for Units
1 and 3.
Within the areas
examined,
no violation or deviation
was identified.
Followup on Generic Letter 88-01,
Supplement
1
(92701)
Supplement
1 provided licensee with acceptable
alternative staff
positions to some of those delineate'd
in GL 88-01,
"NRC Position
on
IGSCC in
BWR Austenitic Stainless
Steel
Piping," dated January
25,
1988.
The alternatives
are in regard to the inspection of reactor water
cleanup
system piping outboard of the containment isolation valves; the
leak detection
requirements
pertaining to the operability of leakage
measurement
instruments;
and the frequency of monitoring leakage rates.
The supplement
also provides clarification or guidance
on the staff's
positions regarding the sample
expansion for Category
D welds; the
effect of shrinkages
resulting from weld overlay repairs or stress
improvement
on the piping system
and its supports
and pipe whip
restraints;
and the technical specification
amendments
for
incorporating the inservice inspection
statement
and leak
detection
requirements
as delineated
in GL 88-01,
TVA is presently working on their submittal
Supplement l.
However, the inspector discussed
TVA's proposed positions with the
Project Hanager for IGSCC mitigation activities at Browns Ferry and
obtained
a detail status of the
IGSCC mitigation activities completed
and ongoing for each of the Browns Ferry Units.
Basically,
TVA's
response will be that they intend to meet the original requirements
of
GL 88-01 at this time.
However,
TVA may re-evaluate
and relax
some of
their positions
when the
new standard
technical specifications
are
implemented.
Within the areas
examined,
no violations or deviations
were identified.
Review of Pipe Support Calculations for Unit 2
The review of Unit 2 calculations
during this inspection is due to the
design
problems
on Unit 3 spring supports
reported
in Inspection
Report
Number 50-259,260,296/92-32
as
an Inspector
Followup Item (IFI) "Design
Problems
in Spring Supports".
The inspector
randomly selected
10 pipe
support calculations for review from two systems,
01 - Hain Steam
system
and
70 - Reactor Building Cooling Water system.
The licensee's
General
Design Criteria,
No. BFN-50-C-7107,
Design of Class
1 Seismic
Pipe
and
Tubing Supports,
was used for these
support calculations
since all of
them were within the scope of IEB 79-14 program.
All five of the
supports
selected
from the Hain Steam
system
were spring can supports.
The remaining five supports,
from the Reactor Building Cooling Water
system,
were rigid supports.
The
10 support calculation were partially reviewed
and evaluated for
thoroughness,
clarity, consistency,
and accuracy.
The review included
checks to see that the applied loads
used
were taken from the latest
stress
calculation,
as well as spring design,
member size,
weld sizes
and symbols,
and standard
component capacities
and settings.
In general,
the design calculations
were acceptable,
except
as noted in
the "Comments,"
below.
The following table
shows the support
calculations
which were partially reviewed
by the inspector.
~5N
Calculation
No.
Rev.
No.
~Sstem
No.
2-47B400S0024
CD-Q2001-881352
1
Ol
Comment:
The spring variability was calculated to be
44%.
An
inadequate
disposition
was found to justify the spring variability of
44% instead of the allowable
25%.
There
was
no record of notification
of the lead civil engineer
found for the calculation.'
Calculation
No.
Rev.
No.
~Sstem
Ne.
2-47B400S0016
CD-Q2001-881384
1
01
Comments:
The spring variability was 31.5%.
There
was
no record of
identification to the lead civil engineer
found in the calculation.
2-47B400S0012
CD-Q2001-882154
2
Ol
Comments:
The spring variability was 36.4%.
There
was
no record of
identification to the lead civil engineer
found in the calculation.
2-47B400S0020
CD-02001-881417
1
01
Comments:
The spring variability was
32%.
There
was
no record of
identification to the lead civil engineer
found in the calculation.
2-47B400S0027
CD-Q2001-882373
0
01
2-47B464H0035
CD-02070-881902
3
70
2-47B464H0029
CD-02070-881980
0
70
Comments:
The latest stress
loads were not incorporated
in the support
qualification.
The angle steel
was also not checked for laterally
unbraced
length per the requirements
of Section 1.4.2. 12 of General
Design Criteria BFN-50-C-7107.
2-47B464S0119
CD-Q2070-881996
2
70
2-47B464R0236
CD-Q2070-883148
1
70
Comments:
The latest stress
loads
were lower than the design loads.
The loads were not evaluated
and/or the results of evaluation for the
impact of the
new load change
documented
in the calculation;
which
should
have
happened
even though the
new loads were lower.
2-47B464S0210
CD-Q2070-883133
2
70
Comments:
The angle steel
was not checked for laterally unbraced
length per the requirements
of Section 1.4.2. 12 of General
Design
Criteria BFN-50-C-7107.
Supports
No. 2-47B400S0024,...S0016,
...S0012,
and ...S0020
had spring
variabilities over the
25 percent
allowed by Section 1.4.4. 1 of General
Design Criteria BFN-SO-C-7107,
Rev.
5.
For spring variabilities over 25
percent Section 1.2.2 of the General
Design Criteria required that
any
conflicts or variances
shall
be identified to the
Lead Civil Engineer
(TVA) before further action is taken
by the designers.
The four support
calculations listed above did not contain
any records of notification to
the
Lead Civil Engineer.
The calculations
showed that dispositions of
the variances
were from the Bechtel
pipe support design
group to the
Bechtel
pipe stress
analysis
group.
Section 1.4.4. 1 of the General
Design Criteria was revised in Rev.
6 to remove the
Engineer responsibility for the variances
and only require
an approval
from the stress
analyst.
k
i
The inspector determined that the disposition of the
44% spring
variability, for Support
No. 2-47B400S0024,
provided
by the Bechtel
stress
analyst,
was inadequate.
The disposition did not consider
an
evaluation of the effect of the variability on the pipe stresses
or
adjacent
support load increases.
The failure to follow the procedural
requirement to inform the
Civil Engineer
about the excessive
spring variabilities,
and the failure
to consider the effect of the excessive variability on the pipe stresses
and adjacent
support loads is considered
to be
a violation against
Appendix B, Criterion V.
Since the licensee
has recently
revised the General
Design Criteria BFN-50-C-7107 to eliminate the
requirement to notify the
Lead Civil Engineer,
and took immediate
corrective action to re-disposition
the
44% spring variability of
Support
No. 2-47B400S0024,
the problems
were considered
to have minor
safety significance
and were not cited because
the criteria specified in
Section VII.B(l) of the
NRC enforcement policy were satisfied.
This
item will be identified as non-cited violation (NCV) 50-260/92-38-02,
"Failure to Properly Identify Support Spring
Can Variability".
The inspector
found that Support
No. 2-47B464H0029
was not based
on the
loads
from latest stress
calculation
No. CD-(2070-880983,
Rev.2,
dated
October 27,
1989.
The loads
used in the support calculation
were from
Rev. 1 of the stress
calculation, while the latest
loads
(from Rev.2)
were increased
about
43 percent
from the original design loads.
In
addition, the angle steel
in this support
was not checked for laterally
unbraced
length per the requirement of Section 1.4.2. 12 of the General
Design Criteria BFN-50-C-7107.
The inspector also found that Support
No.
2-47B464R0236 did not contain documentation that is was evaluated
for the impact of new load changes,
even the
new loads were lower.
The
angle steel
on Support
No.
2-47B464S210
was also not checked for the
laterally unbraced
length per the requirement of Section 1.4.2. 12 of
General
Design Criteria BFN-50-C-7107.
(The licensee's
engineers
stated
that because
the angle size
was 4"
X 4"
X 1/4" and the laterally
unbraced
length was only 18", the design engineers
judged that the
allowable bending stress
would be 0.6
X yield stress
as
a normal design
condition without stress
reduction.
It might be true that the unbraced
length was relatively short,
but it was
an improper design practice to
estimate
the allowable bending stress
without checking the unbraced
length.)
10 CFR Part 50, Appendix B, Criterion III, Design Control requires that
design
changes
shall
be subject to design control measures
commensurate
with those applied to the original design.
TVA Nuclear Engineering
Procedure
NEP-3. 1, Attachment 4,
Page
1 of
1 states
that design inputs,
including information such
as loads,
temperature ... shall
be ...
current,
referenced,
and applied.
TVA 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 calculation for Support
No. 2-47B464H0029
was not
revised to reflect the latest stress
loads which were
43 percent higher
than the design loads.
This Item is identified as Violation 50-
10
260,296/92-38-01,
"Inadequate
Design Controls for Pipe Support
Calculations".
TVA stated that
18 of 38 pipe support calculations
contained
in stress
calculation
No. Nl-270-2R were reviewed
and they had found that this
appeared
to be
an isolated
case.
Within the area
examined,
no deviations
were identified.
Actions on Previous
Inspection
Findings
(92701)
(Open) Inspector
Followup Item (IFI) 50-259,260,296/92-32-01,
"Design
Problems
in Spring Supports"
This IFI involved four concerns
related to spring can design.
The
concerns
were:
two cold-loads in system
068,
inadequate
spring cold-load
setting in a Torus piping system,
two design criteria for Torus
and
other piping systems,
and the Torus piping system not included in the
IEB 79-14 program.
The inspectors
discussed
the matters with the
licensee's
engineers
and reviewed the information provided.
The
licensee
agreed to revise the pipe support calculations
in system
068 to
get one cold-load from the normal operation condition contained
in the
stress
calculation performed
by General
Electric Company
(GE) for the
spring cold setting.
The calculations will be reviewed during the
future inspection.
The licensee
explained their position
on the remaining concerns
as
follows:
- It is true that two different design criteria exist for the Torus
and
other piping systems
because
they were developed
at different times.
The licensee
does not plan to combine the two different design criteria
into one standard
design criteria since it may require more modification
work if the design of all Torus piping systems
are
based
on one standard
design criteria.
The Inspectors will evaluate
the licensee position, in detail, during
a
future inspection.
Within the areas
examined,
no violations or deviations
were identified.
Exit Interview
The inspection
scope
and results
were summarized
on November 6,
1992,
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.
With regards
to Violation No. 50-260,296/92-38-01,
"Inadequate
Design
Controls for Pipe Support Calculations",
the
TVA Site Vice President
stated
he did not agree that this item should
be
a violation since the
mistake did not increase
the load calculations for the support
and,
unless
he did not fully understand
the problem, this was
a isolated
example.
The inspectors
inquired that, if the inspection
continued
after the exit meeting could the licensee
determine that this was
an
isolated
example?
The licensee's
staff stated that the issue could not
be resolved that day.
The inspectors
informed the licensee
the reported
item was
a violation and that further discussion
would only attempt to
establish
the severity level which is not determined
by the inspectors.
Therefore,
the Vice President
comments
would be discussed
with the
appropriate
NRC management
before severity levels are assigned.
(Open) Violation No. 50-260,296/92-38-01,
"Inadequate
Design Controls
for Pipe Support Calculations",
paragraph
7
(Open)
NCV No. 50-260/92-38-02,
"Failure to Properly Identify Support
Spring
Can Variability", paragraph
7
(Open)
URI No.50-296/92-38-03,
"Evaluation of Weld Conditions",
paragraph
3
and Initialisms
ASHE
BS,PV
E/C
GL
IGSCC-
IN
TS
American Society For Hechanical
Engineers
Boiling Water Reactor
Boiler and Pressure
Vessel
Code
Erosion
and Corrosion
Electric Power Research
Institute
System
General
Electric
Generic Letter
Intergranular Stress
Corrosion Cracking
NRC Information Notice
Inservice Inspection
Nondestructive
Examination
Nuclear Reactor Regulations
Quality Assurance
Reactor
Water Cleanup
System
Technical Specifications
Unresolved
Item
Ultrasonic Testing
Visual Testing