ML17290A577
| ML17290A577 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 08/13/1993 |
| From: | Miller L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17290A575 | List: |
| References | |
| 50-397-93-25, NUDOCS 9308230260 | |
| Download: ML17290A577 (19) | |
See also: IR 05000397/1993025
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION V
~Re ort Ro.:
50-397/93-25
ocket No.:
50-397
License No.:
Licensee:
Washington Public Power Supply System
(WPPSS)
~li:
Il tilt
g
I
Pl t,gltt
Benton County, Washington
Ins ection Conducted:
July 12-23,
1993
~lt
~ddtt
D. Acker, Reactor Inspector
H. Roy ck, Reactor
spector
D.
C
y, Pr
ct
spector
er,
r.,
C
e
eactor
a ety
rane
Ins ection
Summar
Ins ection durin
the
eriod of Jul
12-23
Re ort No. 50-397 93-25
I
t d:
During this routine announced
inspection the inspectors
reviewed selected
design
changes
and previously identified items.
Inspection
Procedures
37700,
"Design Changes,"
and 92701,
"Followup," were used for this inspection.
Safet
Issues
Mana ement
S stem
SINS
Ite
None
Results:
General
Conclusions
and
S ecific Findin s:
The design
changes
reviewed were technically adequate.
The licensee
had not updated preventive maintenance
instructions to include
new safety related
equipment installed
by three completed
design
changes.
Procedures for signing verification of completion of design
change
steps did
not always include appropriate
signature
blocks.
9308230260
930813
- DOCK 05000397
8
~
~
~
Si nificant Safet
Batters:
None
Summar
o
Violation or Deviations:
One violation of 10 CFR 50, Appendix B, Criterion V, was identified in Section
2.1.1.
0 en Items
Summar
Three followup items were closed.
One enforcement
item was opened.
,
Persons
Contacted
DETAILS
Washin ton Public Power
Su
1
S stem
- ¹R. Barbee,
Manager,
System Engineering
¹H. Flasch, Director of Engineering
¹J.
Gearhart,
Director, guality Assurance
¹P.
Harness,
Manager,
Mechanical
Design
- L. Harrold, Manager,
Maintenance
- ¹R. Koeings,
Manager,
Design Engineering
- ¹R. Mathews,
Manager,
Electrical/ISC,
Design Engineering
¹T. Messersmith,
Manager,
Maintenance
Support
- ¹L. Oxsen,
Deputy Managing Director
¹J. Parrish, Assistant
Managing Director
¹K. Pisarcik,
Licensing Aide
¹J.
Rhoads,
Manager,
Operating
Events Analysis
and Resolution
- H. Rice, Plant Support Engineer
- ¹J. Sorensen,
Manager,
Regulatory Compliance
- ¹J. Swailes,
Plant Manager
- ¹D. Swank,
Licensing Engineer
- ¹S. Washington,
Manager,
Nuclear Safety Engineering
¹R. Webring, Manager,,Technical
Division
US Nuclear
Re ulator
Commission
2.
- D. Acker, Reactor Inspector
- W. Ang, Engineering Section Chief
¹R. Barr, Senior Resident
Inspector
¹K. Johnston,
Project Inspector
¹D. Proulx, Resident
Inspector
¹H. Royack,
Reactor Inspector
- Denotes those attending the exit meeting
on July 15,
1993
¹Denotes
those attending the exit meeting
on July 22,
1993
The inspectors
also held discussions
with other licensee
personnel
during the course of the inspection.
Desi
n Control
37700
2.1
Desi
n Chan
es
The inspectors
reviewed six basic design
changes
(BDCs) to safety
related
equipment
which the licensee
had determined to not require prior
NRC approval.
The inspectors
reviewed the
BDCs for conformance with
Technical Specifications,
10 CFR 50.59, the licensee's
quality assurance
program,
and
10 CFR 50, Appendix B, Criterion III, "Design Control."
The inspectors
reviewed the following BDCs:
~
BDC 88-0442-0A,
"High Pressure
(HPCS) Solenoid
and Air
Pressure
Control Valve Replacement"
BDC 93-0021-0A,
"SM-7 and
SM-8 Relay Coordination"
BDC 93-0024-0A,
"Voltage Regulator Relay Configuration
Modification for DG2"
2.1.1
BDC 89-0218-0A,
"High Pressure
(HPCS) Test Return Line
Restricting Orifice"
BDC 90-0288-0A, "Critical Switchgear
Normal Cooling"
BDC 93-0082-0A,
"Reactor
Core Isolation Cooling System
(RCIC)/Containment Isolation Interface"
The inspectors
chose
BDC 88-0442-OA for review because it had
been
entirely completed.
BDCs 93-0021-OA and 93-0024-OA were chosen
because
they had been recently installed
and declared
in 1993.
BDC 93-
0082-OA was chosen
because it was
an "Urgent Modification." This allowed
the inspectors to evaluate
both the entire design
process
and recent
design work.
The inspectors
evaluated
each
BDC for approval authority, procedure
control, proper testing criteria, proper licensee
updating of operating
procedures
and training,
as built drawing control, proper safety
evaluations,
proper licensee
updating of maintenance
procedures,
and
control
and update of the Updated Final Safety Analysis Report
(UFSAR)
The inspectors
concluded that the
BDCs reviewed
met the review criteria
except for one violation for failure to update preventive maintenance
instructions for three newly installed safety related
components.
Preventive
Maintenance
The licensee
used
scheduled
maintenance
system
(SMS) Data Input Sheets
to add
new equipment to their routine preventive maintenance
program per
licensee
procedure
Plant Procedure
Manual
(PPM) 10.1.5.
Plant Procedure
Manual
(PPM) 1.4.1, Revision 14, "Plant Modifications,"
Paragraph
5.4,
Step
1 required that the assigned
project engineer
initiate and coordinate
a Plant Modification Record
(PMR) Package
Checklist for design
changes.
The
PMR package checklist included
a
block to check if SMS data input sheets
were required.
The inspectors
reviewed
PMR package checklists for BDC 93-0021-OA and
noted that the
SMS data input sheets
were not required.
However, the
inspectors
determined that
BDC 93-0021-OA was based
on another recent
modification, 91-0222-0A,
"DG-2 Field Cutoff Relays," which did require
SHS data input sheets.
The inspectors
were concerned that the
PHR
package checklist for BDC 93-0021-OA had not been appropriately
completed
as required
by
PPH 1.4.1.
PPM 1.4.1,
paragraph
5.5, step
20
required that the assigned
project engineer
sign in the
PHR Package
Checklist that the
SHS Data Input Sheets
were completed.
The inspectors
noted that the
PHR Package
Checklist did not contain
a signature
block
for this signature.
The inspectors
reviewed the
PMR checklists for BDCs 88-0442-0A,
93-0024-
OA and 91-0222-OA.
The inspectors
noted that the licensee's
pl oject
engineer
had appropriately
checked that
SHS data input sheets
were
required.
However, the inspectors
were unable to locate
any
SHS data
input sheets
or signatures
of completion of the
SMS data input sheets
for these three
BDCs.
Because
SHS data input sheets
had not been
initiated,
new safety related
equipment including emergency diesel
generator
EDG pressure
regulating valves,
start
sequence
timing relays,
and
4160 volt power coordination relays
had not been included in the licensee's
preventive maintenance
program.
In response
to the inspector's
concern the licensee initiated
SHS data
input sheets for these three
BDCs.
The licensee
noted that for BDC 88-
0442-OA they intended to include the
new equipment in preventive
maintenance
using
a procedure
change.
Failure to complete
and sign for
these
SMS data input sheets
is
a violation of 10 CFR 50, Appendix B,
Criterion
V (Violation 50-397/93-25-01).
2. 1.2
U dated Final Safet
Anal sis
Re ort
I
Engineering Instruction (EI) 2.8, Revision 9, "Generating Facility
Design
Change
Process,"
required that
a design safety analysis
be
included
as part of a BDC.
The inspectors
noted that the design safety
analysis for BDC 93-0024-OA indicated that the design
change affected
Chapter
15 of the Updated Final Safety Analysis Report
(UFSAR) and that
a Safety Analysis Report
Change Notice
(SCN) was required to be
initiated.
However, the
PHR package checklist for BDC 93-0024-OA was
checked to indicate that
an
SCN was not required.
In response
to the inspector's
observation,
the licensee
reviewed
93-0024-OA and concluded that
an
SCN was not required.
The licensee
noted that
BDC 93-0024-OA only changed
a
UFSAR drawing, which was
planned to be updated in the next
UFSAR update.
The inspectors
reviewed
the licensee's
records
and identified that the drawing in question
was
listed for UFSAR updating.
The inspectors
also reviewed the design
change
and concurred with the licensee that
no
UFSAR text changes
were
required.
The inspectors
considered that the difference
between the design
analysis
and the
PHR Package
Checklist should have
been resolved
by the
licensee
as part of their design review process.
The licensee
concurred.
4
2. 1.3
Verification of Com leted Actions
During review of the
BDCs, the inspectors
noted
a number of examples
where tables
and checklists
associated
with verification of completion
of a
BDC did not match the associated
instructions.
Examples of these
mismatches
are listed below.
~
As noted in Section 2.1.1 of this report,
PPM 1.4.1,
Paragraph
5.5,
Step
20 required that the assigned
project engineer
sign in
the
PHR package checklist that the
SHS data input sheets
were
completed.
The
PHR package checklist did not contain
a signature
block for this signature.
PPM 1.4. 1,
Paragraph
5.5,
Step
5 required that the assigned
project engineer identify plant procedures
affected
by a
BDC,
initiate actions to update
these
procedures,
and sign the
appropriate
blank on the
PHR package checklist.
The
PHR package
checklist did not contain
a signature
block for this signature.
PPM 1.4. 1, Paragraph
5.6,
Step
1 required that the assigned
project engineer review the entire
PHR package
including all
appropriate checklists
and then sign
and date the
PHR.
The only
signature
space
on the
PHR for the project engineer
was titled,
"Critical Documents
Updated."
The inspectors
discussed .the "Critical Documents
Updated," signature
space with several
project engineers
and got different opinions
as to
what this signature required.
PPM 1.4.1 did not define Critical
Documents.
For the
BDCs reviewed,
the inspectors
did not identify any resulting
problems
due to the above procedural
deficiencies,
except
as noted in
Section 2.1.1 of this report.
2. 1.4
Discussion
and Conclusions
The inspectors
determined that the
BDCs reviewed were technically
adequate.
The inspectors
determined that the licensee
had adequate
checklists for
identifying actions required to be taken
as part of a design
change.
However, the inspectors identified examples
where the procedures
and
checklists for verification of completion of the required actions
were
mismatched.
In general,
the inspectors
also found the licensee's
verification of completion of actions
was not as well documented
as
their identification of required actions.
The inspectors
were concerned
that the procedure
format deficiencies
created
the potential for
procedure violations since
no blank spaces
existed to highlight
incomplete 'actions.
The inspectors
observed that the procedures for verification of
e
2.2
completion of BDCs could contribute to procedure
compliance problems.
In response
to the'nspector's
concerns,
the licensee
committed to
review and update
as necessary
the procedures
or checklists to clearly
indicate what actions
and verifications were required.
Tem orar
Modifications
3.
The inspectors
reviewed two temporary modifications for program
controls,
procedure details,
approval responsibility,
formal records of
the changes,
independent verifications of the changes,
functional
testing, periodic licensee
review and adequacy of the design.
The following temporary modification requests
(THRs) were reviewed:
~
TMR 92-024: Disconnect
and remove
bad nitrogen
system temperature
switches
and install pipe plugs
~
TMR 92-062:
Remove covers to safety related microprocessor
and
relay drawers
The inspectors
concluded that the THRs reviewed
had adequate
program
control, proper level of procedural detail to complete the task, correct
level of approval,
records of changes,
independent verification of
changes,
post installation testing,
and periodic licensee
review of
design
adequacy.
The inspectors
determined that the licensee's
quarterly report
on
outstanding
THRs did not accurately reflect the actual installation date
for THRs installed prior to March 1992.
In March 1992 the licensee
changed
the
THR system,
and in converting older THRs to the new system,
used the date of the conversion
as the installation date in quarterly
reports in lieu of the actual installation date.
The inspectors
considered that
use of the conversion
date could mislead
management
on
the effectiveness
of actions to remove
and close
THRs.
The licensee
acknowledged
the inspectors'oncerns.
Followu
92701
3.1
Closed
Followu
Item 50-397 92-25-07:
Reactor
Core Isolation Coolin
Turbine Lube Oil
Sam les
Ori inal
NRC 0 en Ite
NRC inspectors
had previously reviewed reactor core isolation cooling
(RCIC) turbine lube oil samples
and
had noted that the particle counts
were higher than "Terry Turbine Controls Guide," NP-6909
recommended
maximums.
Licensee
RCIC turbine lube oil samples for November 3,
1991,
February
25,
1992,
and April 20,
1992, indicated particle counts of
400,000,
400,000,
and 238,800 particles, respectively,
in the
5 15
micron range for a 100 milliliter lube oil sample.
The inspector
noted
0
W
that Terry Turbine maximum recommended
standards
for 5 - 10 and
15 - 25
micron was 24,000 particles
and
5360 particles,
respectively, for a
100
millilitersample.
The licensee
had not taken
any corrective actions for the higher than
normal lube oil particulate indications since
changes
were being performed at greater
than required frequencies,
and
since turbine vibration, lube oil differential pressure,
and bearing
temperatures
were not increasing.
icensee's
Actions in Res
onse to the 0 en
Item'oblem
Evaluat
on
Re uest
The licensee
issued
problem evaluation request
(PER) 292-986,
"RCIC Lube
Oil," to evaluate
the higher than normal particulate
count in the
PER 292-986
recommended
the following:
~
Contact Terry Turbine to request their recommendation
on cleaning
up the lube oil and determine if the turbine was seriously
affected
by the high particulate count.
Perform the
SMS task to obtain
a current oil sample instead of
waiting until the next scheduled oil sampling period.
~
If the current oil sample particulate
count was high, develop
a
plan to clean
up the oil system.
Licensee Actions
The licensee
contacted Terry Turbine Controls to determine if the higher
than
recommended
lube oil particulate
count would have
an effect on the
RCIC Terry Turbine or its controls.
The licensee
documented their
conversations
in records of telephone
conversations
dated August 20,
1992.
The records of conversation
concluded that lube oil particulate
counts would not have
an adverse affect on the control or operation of
the tur bine since internal filtering systems of 20 to 25 microns were
installed.
The licensee
sampled the
RCIC turbine lube oil and additionally sampled
an old and
a new drum of Mobil Oil Vaprotec oil.
Vaprotec oil is the
type of lube oil used in the
RCIC turbine.
The old lube oil drum sample
was taken from a drum used to fill the
RCIC turbine lube oil reservoir.
The lube oil analysis
concluded that the turbine oil sample
had
particulate count levels of 269,685 for 5 - 15 micron sized particles
and 2,003 for 15 - 25 micron sized particles.
The
5 - 15 micron sized
particle count was higher than the recommended,
however,
15 - 25 micron
sized particle count was lower than the recommended.
The lube oil
samples
taken from the new and old lube oil drums also
had particulate
count levels of 898,790
and 281,335 for particle sizes of 5 15'
microns,
and 301,051
and 63,645 for particle sizes of 15 - 25 microns,
respectively.
Therefore the licensee
had concluded that the particulate
were being introduced into the
RCIC turbine from the lube oil.
The licensee
had introduced
a program to pre-filter lube oil prior to
installation into equipment.
The filtering process
had filtered the
lube oil when it was received on-site
and again
when it was placed into
the unit.
Therefore, after several
RCIC turbine cycles of lube oil
system
changes
and flushing the particulate
count level could be reduced
to recommended
particulate
count levels.
The licensee
stated that they were evaluating
changing the type of lube
oil used in the
RCIC turbine.
Ins ecto s'ctions
Durin
the Present
Ins ection
The inspectors
reviewed licensee
records of conversation
between
the
licensee
and the
RCIC turbine control supplier.
The inspectors
reviewed
licensee
documentation of RCIC turbine lube oil samples
and the
new and
old drum lube oil samples.
The inspectors
determined that both the
new
and the old lube oil drum samples
had
a higher particulate count level
than the lube oil that was pre-filtered
and used in the
RCIC turbine.
The inspectors
reviewed licensee letters to lube oil suppliers
confirming the acceptability of alternate
lube oils for the
turbine.
Discussion
and Conclusion
The inspectors
concluded that the licensee
had verified with the
turbine control supplier that the
above
recommended
RCIC turbine lube
oil particulate levels did not affect the turbine controls or operation.
The inspectors
concluded that higher than
recommended
particulate
had
been introduced into the
RCIC turbine in the lube oil and that the
licensee
had initiated pre-filtering the
RCIC turbine lube oil to reduce
the problem.
The inspectors
also concluded that the licensee
had taken
appropriate
action to determine the cause of and reduce the particulate
count levels in the
This item is closed.
3.2
No violations or deviations of NRC requirements
were identified.
Closed
Followu
Item 50-397 92-26-01:
Pi in
Calculations
Ori inal
NRC 0 en Ite
This follow-up item concerned
the adequacy of licensee
procedural
guidance for overlapping piping calculations
in the absence
of a single
anchor-to-anchor
calculation.
The inspector
had
been
concerned that (I)
licensee
Piping Design Guide,
HES-3 suggested
that two piping restraints
in each of the three orthogonal directions
was sufficient to establish
a
boundary for overlapping calculations,
which was less conservative
than
the guidance in NUREG 1980,
"Dynamic Analysis, of Piping Using the
Structural
Overlap method,"
and (2) that the licensee
might have
applied the less conservative criteria to safety-related
piping analyses.
The licensee
had committed to develop
a plan to determine if any safety
related piping had used the overlap method of analysis
and to review the
Piping Design Guide MES-3 for adequacy.
icensee's
Actions in Res
onse to the 0 en Item
The licensee
revised Design Guide
(MES-3) to be consistent with the
guidance
given in NUREG 1980 for overlap criteria calculational
analysis.
The licensee
sampled
160 safety-related
large bore piping calculations
to determine if the overlap method for piping analysis
was used.
The
160 calculations
sample
was greater than
50K of the large bore piping
calculations in this category.
The licensee identified two calculations
which were performed using the overlap method.
The two calculations
were performed for service water system piping.
The licensee
reviewed
the two service water
system calculations
and found that both
calculations
met the guidance
provided in
NUREG 1980 for overlapping.
Ins ectors'ctions
Durin
the Present
Ins ection
The inspectors
reviewed licensee
Design Guide MES-3, Revision 1,
and
determined that the licensee
had issued
procedure
amendment
92-10
on
October
14,
1992, for the design guide to incorporate
guidance.
The inspectors
determined that the licensee
had
sampled safety-related
piping calculations to determine
which calculations
had used the overlap
method of calculation,
and that two calculations of 160 were found to
have
used the overlap method of analysis.
The inspectors
determined that the methods
used to develop the overlap
model for the two service water piping calculations
met the guidance of
Discussion
and Conclusion
The inspectors
concluded that the licensee
had reviewed
and
appropriately revised
Design
Guide MES-3.
The licensee
had developed
and implemented
a plan to adequately
sample large bore piping stress
calculations to determine if any had
used the overlap method.
The
licensee
had determined that the calculations
which used the overlap
method of calculation
had results
which met the criteria of NUREG 1980.
This item is closed.
No violations or deviations of NRC requirements
were identified.
e
9
Closed
Followu
Item 50-397 92-26-02:
Pi in
Su
ort Calculations
Ori inal
NRC 0 en Ite
This follow-up item concerned
the licensee's
calculational
method for
combining independent
support group responses
on certain
ASME Code Class
1 piping response
spectrum
analyses
as listed in WNP-2 Technical
~Memorandum
1303,
dated August 10,
1983, "Multiple Input Response
Spectrum Analysis Method of Combining Responses
Due to Individual
Support Excitations."
The inspector
had
been
concerned that the
independent
support group responses
in multiple level response
spectra
analyses
(MLRS) were being combined
by the square root of the
sum of the
squares
method
(SRSS),
which was
a less conservative
approach
than the
absolute
summation
method
(ABS) recommended
in NUREG 1061,
Volume 4.
Although the
SRSS method
had been
used
on certain Technical
Memorandum
1303 calculations
at the time of licensing, the inspector
observed that
the licensee
had used the
same
SRSS method for performing calculations
to justify piping snubber reduction for the subject piping.
Since the
SRSS method
was less conservative
than the
ABS method of calculation,
the inspector
was concerned that the results
from the
SRSS method
used
to justify snubber reduction
may not have
been conservative.
"In response 'to the inspector's
concern,
the licensee
had committed to
perform a technical
evaluation of their calculations,
which used the
SRSS method,
considering industry studies
including EPRI report NP-6153,
Seismic Analysis of Multiply Supported
Piping Systems."
Lice see's
Actions 'n Res
onse to the 0 en Ite
The licensee
performed a'technical
evaluation of their piping
calculations,
which used the
SRSS method,
considering industry studies
including EPRI report NP-6153,
and Melding Research
Council Bulletin
352, dated April 1990.
The licensee
stated that future calculations
would continue to limit the
SRSS combination of independent
support group responses
to those
calculations
which were performed in a similar manner at the time of
licensing.
Ins ectors'ctions
Durin
the Present
Ins ection
The inspectors
reviewed the licensee's
evaluation
and noted the
following:
~
The licensee's
approach to combining independent
support group
responses
by SRSS
was consistent with the method
recommended
in
Melding Research
Council Bulletin 352, dated April 1990.
~
EPRI report NP-6153 found that
SRSS combination
between
support
group responses
provides generally conservative
responses
when
compared to test data.
10
The inspector determined that the licensee's
technical
evaluation
demonstrated
that the use of SRSS for combining independent
support
motion responses
was consistent
with EPRI report NP-6153.
The inspector
also determined that the licensee's
use of the
SRSS
method
was generally
conservative for the piping calculations listed in licensee
Technical
Memorandum
1303.
Discussion
and Conclusion
The inspectors
concluded that the licensee
had adequately
performed the
technical
evaluation of their calculations
which used the
SRSS
method
for combining independent
support motion responses.
The inspectors
also
concluded that the evaluation
demonstrated
the
SRSS
method to be
conservative for the piping calculations listed in WNP-2 Technical
Memorandum
1303.
This follow-up item is closed.
No violations or deviations
were noted in the areas
inspected.
~Eit tt
The inspectors
conducted exit meetings
on July 15,
1993,
and July 22,
1993, with members of the licensee staff as indicated in Section
1.
During these meetings,
the inspectors
summarized
the scope of the
inspection activities
and reviewed the inspection findings as described
in this report.
The licensee
acknowledged
the concerns identified in
the report.
The licensee did not identify as proprietary
any of the
information provided to the inspector.
bcc w/enclosure:
P. Johnson
Project Inspector
Resident
Inspector
G.
Cook
B. Faulkenberry
Docket File
bcc w/o enclosure:
H. Smith
J. Zollicoffer
J.
Bianc
DAcker
MRoyack 4 w'Huey
g /ii
93
V/0/93
~/~93
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Project Inspector
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Docket File
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