ML17229B515
| ML17229B515 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 07/05/1996 |
| From: | Kurth M, Christine Lipa, Kenneth Riemer NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17229B513 | List: |
| References | |
| 50-331-96-04, 50-331-96-4, NUDOCS 9607310047 | |
| Download: ML17229B515 (22) | |
See also: IR 05000331/1996004
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION III
Docket No:
License
No:
50-331
Report
No:
50-331/96004
Licensee:
IES Utilities Incorporated
IE Towers,
P. 0.
Box 351
Cedar Rapids,
52406
Facility:
Duane Arnold Energy Center
Dates:
April 20 June 5,
1996
Inspectors:
K. Riemer,
Senior Resident
Inspector
C. Lipa, Resident
Inspector
M. Kurth, Reactor
Engineer
G. Kelly, NRR Project Manager
Approved by:
R.
D. Lanksbury, Chief
Reactor Projects
Branch
2
96073i0047
960705
ADOCK'500033f
8
EXECUTIVE SUMMARY
Duane Arnold Energy Center
NRC Inspection
Report 50-331/96004
This integrated
inspection report included aspects
of licensee
operations,
engineering,
maintenance,
and plant support.
The report covers
a 6-week
period of 'resident
inspection;
in addition, it includes the results of
announced
inspections
by the
NRR project manager
and
a regional reactor
inspector.
~0evations
The inspectors
identified that Operation's
had not recognized,
prior to
authorizing
a maintenance activity,
a required entry into a limiting
condition for operation.
This was
a violation (Section Hl.3).
The inspectors
identified an increasing
trend in offgas pretreat
radiation release
rate that
had not been identified by the operators.
(Sections
01.3
and Rl.l)
Operator
response
to materiel condition problems
encountered
during the
period,
such
as
a reactor recirculation motor generator trip and
scoop
tube lockups,
continued to be appropriate
(Section H2).
Maintenance
One station activity undertaken
in response
to the Hay 16,
1996,
unexpected trip of the "B" reactor recirculation motor generator
(RRHG)
set
and control of monitoring equipment installed
because
of earlier
trips, were weak.
This resulted in the loss of potentially important
data necessary
to determine
the root cause of the trip (Section H1.4).
Chronic problems with the well water chlorination system continued to
challenge plant staff because
of resultant
higher than 'normal drywell
temperatures
due to fouling (Section
M2).
En ineerin
Engineering
support to plant operations
during this inspection period
was mixed.
Short term resolution of materiel condition issues
was
appropriate.
However, actions
taken to date in response
to long term
concerns
such
as the
RRHG trips and well water chlorination issues
have
not resolved
the items,
nor have they prevented further challenges
to
plant staff.
Engineering participation in the activities associated
with the "8" reactor recirculation motor generator
set
was weak and
contributed to the loss of potentially important data
(Section
M1.4 and
H2.1).
The inspector's
review of the containment
hardened
vent modification
identified only minor discrepancies,
which the licensee
properly
resolved
(Section El.l).
Several
UFSAR discrepancies
were identified during the Spent
Fuel
Pool
Licensing Basis
Review (Section E2.1).
Plant
Su
ort
A potentially negative trend in the off-gas radiation release
rate
was
not effectively communicated
by Chemistry through the site organization
until the inspectors
brought the matter to the licensee's
attention
(Section Rl).
Se f Assessment
and
ualit
Ve if'cation
Self-assessment
activities,
such
as Operations
Committee
and Action
Request
screening
meetings
were considered
effective (Section 07).
The
gA surveillance of the containment
hardened
pipe vent modification
was considered
thorough
and detailed
(Section El.l).
Re ort Detail s
Summar
of Plant Status
The plant began this inspection period at
100 percent
power.
There
was
a
routine downpower for turbine valve testing
on May 11.
From May 16 until
May 23, the plant was in single loop operations
and reactor
power was
approximately
45 percent following the trip of the "B" reactor recirculation
motor generator
(RRMG) set.
Following recovery of "B" RRMG on May 24," 1996,
the plant operated
near
100 percent
power for the remainder of the inspection
period.
I. 0 erations
01
Conduct of Operations
Ol.l
General
Comments
a.
Ins ect
o
Sco
e
71707
Using Inspection
Procedure
71707, the inspectors
conducted
frequent
reviews of ongoing plant, operations.
This included control
room
observations
and plant tours.
The inspectors
noted that, the conduct of
operations
was professional
and safety conscious.
Observations
indicated that the operations staff was knowledgeable of plant
conditions,
responded
promptly and properly to alarms,
adhered to
procedures
and applicable administrative controls,
performed through
turnovers,'nd that proper control
room staffing levels existed.
01.2
Failure To Reco nize
LCO Entr
b.
Observations
and Findin s
As discussed
in Section M1.3, the inspectors identified during
independent verification of operator actions for an ongoing maintenance
activity that the operators
had failed to recognize, prior to
authorizing maintenance
on
a drywell pressure
instrument, that entry
into a limiting condition for operation
(LCO) was necessary.
The
inspectors, were concerned that operation's
involvement in planning
and
review of this maintenance activity was weak.
01.3
0 erators
Unaware
o
Increasin
Trend in Off as Pretreat
Radiation
Release
Rate
b.
Observations
and F'in
s
As discussed
in Section Rl.l, control
room operators
were not aware of
an increasing
trend in the offgas pretreat radiation release
rate until
the inspectors
brought it to their attention'.
The increasing trend was
potentially an early indicator of fuel leakage
problems.
The operator's
il
I.
01.4
immediate
response
was to contact the chemistry department
to assist
in
resolution of this issue.
The inspectors
were concerned that
a
potential early indication of a fuel problem was not noted
by the
operators
in a timely manner.
A
ro riate
0 erator
Res
onse to Reactor Recirculation
Pum
Tri
a
~
Ins ection
Sco
e
71707
b.
On Hay 16,
1996, the "B" reactor recirculation motor generator
(RRHG)
set unexpectedly tripped
as discussed
in Sections
H1.4 and H2. lb.
The
inspectors
independently verified that appropriate
actions
were taken
by
reviewing strip charts,
surveillance
requirements,
and technical
specification requirements.
Observations
and Findin
s
01.5.
~
~
07
07.1
The operators
responded
appropriately to the event
and successfully
maintained
the plant in single loop operations.
The operators
also
correctly implemented the increased
surveillance
requirements
and
administrative controls necessary
to maintain the plant in single-loop
operation while trouble-shooting
and maintenance activities were
performed
on the "B" RRHG set.
On Hay 23,
1996, the inspectors
observed
the operators
place the "8" loop of RR back in service
and return the
unit to full power.
Conclusions
on Conduct of 0 erations
The inspectors
determined that, with the exception of the maintenance
activity on
a drywell pressure
instrument,
operator
cognizance
and
oversight of maintenance activities were appropriate for the tasks
performed.
The inspectors
also concluded that operators
performed well
in response
to the
RRHG set trip and during associated
troubleshooting
and maintenance activities,
and during recovery of the idle loop and
return of the unit to full power.
With the exception of the increasing
trend in offgas pre-treat radiation release
rate levels,
operator
panel
awareness
was thorough.
guality Assurance
in Operations
Licensee
Self-Assessment
Activities
40500
b.
Observations
and Findin
s
During the inspection period, the inspectors
reviewed multiple licensee
self-assessment
activities, including:
Routine Operations
Committee Meeting
Routine Action Request
Screening
Meetings
Special
Operations
Committee meeting
on Hay 17 to review
recommendation
to operate
at higher power level in single loop
operations,
as allowed
by TS.
Special
Operations
Committee meeting'n
Hay 22 to discuss
restoration of the "B" RRHG set
and return to full power
operations.
f
The inspectors
observed that there
was active participation at the
meetings.
c.
Conclusions
on License Self-Assessment
Activities
The inspectors
concluded that the self-assessment
activities observed
were effective.
08
Niscellaneous
Operations
Issues
(92700)
08.1
Closed
Licensee
Event
Re ort
LER
50-331
94003
Revision 0:
Failure
to Establish
Secondary
Containment
During Routine Maintenance.
The
licensee identified the cause to be incomplete
implementation of a TS
amendment.
The issue
was determined to be
a non-cited-violation
as
discussed
in inspection report (IR) 50-331/94002.
The inspectors
reviewed the completion of corrective actions listed in the
LER and
had
no concerns.
This item is closed.
08.2
Closed
LER 50-331
94006
Revisions
0 and
1:
Reactor
Water Cleanup
Isolation
Due to Incomplete Valve Closure
Caused
by Position Indicator
Obstruction.
This issue
was originally discussed
in IR 50-331/94008.
The inspectors
reviewed the completion of the four'orrective actions
discussed
in the
LER.
The corrective actions
were considered
appropriate
to prevent recurrence.
This item is closed.
II. Maintenance
Nl
Conduct of Naintenance
Ml. 1
General
Comments
a.
Ins ection
Sco
e
62703
61726
The inspectors
observed
and/or reviewed portions of the following work
and testing activities:
Drywell pressure
recorder maintenance
Deluge system for standby transformer
Turbine control valve
RPT logic and
RPS instrument functional
test
Daily instrument
checks
Functional/calibration of reactor water level instrumentation
system
boron concentration test
Standby diesel
generators
monthly operability test
"D" river water supply
pump motor replacement
Reactor recirculation motor generator
set troubleshooting
M1.2
Instrument
Used for TS Readin
not in Calibration
Pro
ram
b.
Observations
and Findin
s
On April 25,
1996, the inspectors
identified that level instrument
LI 3413,
used to verify fuel pool level
as required
by TS 4.9.C was not
in a calibration program.
Surveillance
Test Procedure
STP42A001,
"Daily
Instrument
Checks,"
Revision
113, required that fuel pool level
be
documented daily using LI 3413,
however,
the instrument
had not been
calibrated
since it was installed in 1990.
The licensee
was able to
show that the design of the instrument
was
such that periodic
calibration
was not necessary.
Upon further review, the inspectors
considered this to be
an isolated
case;
however,
the inspectors
were
concerned that nothing in the modification process specifically required
assigning
a calibration frequency to new instruments that would be used
for TS surveillances.
The licensee
promptly documented
the condition
on
AR 96-0759.
Pending
further review by the
NRC and the results of the licensee's
evaluation,
this is considered
an Unresolved
Item (URI 50-331/96004-01).
Failure to Enter
LCO Durin
Maintenance Activit
Observations
and Findin
s
Prior to the removal of drywell pressure
recorder
instrument
PR4398A for
maintenance
in accordance
with CHAR A26453
on Hay 15,
1996, the
Operations Shift Supervisor failed to realize that the channel
would be
which required entry into a 30-day limiting condition for
operation
(LCO) in accordance
with TS Table 3.2-H.
The inspectors
identified this discrepancy
on Hay
17 and brought it to the
OSS'ttention.
Technical Specifications
Table 3.2-H requires
a minimum of two channels
operable for the Drywell Pressure
Monitor (0-250 psig).
Technical
Specification Action Statement
93 required that if the number of
channels
was reduced to one channel,
then entry into a 30-day
LCO was required provided
a redundant
channel,
specified in Table 3.2-F,
was operable.
The redundant
channel
was operable at the time.
Technical Specification 6.8. 1 specified that written procedures
covering
areas
such
as corrective maintenance
be implemented.
Maintenance
Directive
(HD) 020,
"Maintenance
Planning," Revision 25, specified that
the
OSS was responsible for determining the effect on the plant and
any
other
requirements"
or special
conditions that were required for the
maintenance
to occur.
Failure of the
OSS to identify that there
was
a
TS
LCO associated
with the maintenance
was
a violation (50-331/96004-02)
of TS 6.8. 1.
The inspectors
were concerned
that planning
and review
prior to maintenance
were not adequate,
which was due to personnel
error
and inadequate
attention to detail.
Poor Troubleshootin
Activities Followin
Tri
at "B" Reactor
Recirculation Motor Generator
RRMG
Set
Observations
and Findin
s
Following an earlier trip of the "B" RRMG set,
the licensee installed
monitoring equipment
on the unit in order.-to obtain data necessary
to
determine
the root cause
for, the repeat trips experienced
on the
machine.
Subsequent
to the Hay 16,
1996, trip of the "B" RRHG set,
engineering
and maintenance
personnel
attempted to retrieve data
received after the event.
Their attempts
were unsuccessful;
none of the
7
H1.5
monitored parameters
on the
"B",RRHG set
had
any data available.
The
parameters
measured
on the "A" RRHG set were all available
and indicated
normal,
expected
values.
Further investigation
by the licensee
determined that the data
was unavailable
on the'B" side
because
the
trigger points
had not been
armed
and turned
on following the trip of
the "B" RRHG set
on January
17,
1996.
Potentially important data
necessary
for root cause determination
was lost because
licensee
personnel
failed to correctly utilize the installed monitoring
equipment.
Additional licensee
troubleshooting efforts focused
on
a transformer in
the voltage regulator circuit which was noted to have
blown fuses'.
Engineering
and maintenance
personnel
removed the component
from the
plant and attempted to bench test it in the, shop.
Improper electrical
troubleshooting'methods
were applied
and the transformer
catastrophically failed.
Because of human error, the only piece of hard
physical
evidence
was destroyed;
again, potentially important data
necessary
for root cause determination
was lost.
The inspectors will continue to monitor the licensee's
investigation of
the above
personnel
errors
and subsequent
corrective actions.
This will,
be tracked
as
URI 50-331/96004-03.
Conclusions
on Conduct of Maintenance
H2
H2.1
The inspectors
noted that most maintenance
activities during the period
were completed thoroughly and professionally.
However, the inspectors
were concerned
with personnel
errors during troubleshooting of the trip
of the "B"'RHG set
and inadequate
attention to detail in planning for
the drywell pressure
monitor maintenance,
discussed
above.
Naintenance
and Nateriel Condition of Facilities and Equipment
Plant Material Condition
b.
Observations
and Findin
s
Plant materiel condition was acceptable.
The inspectors
noted that
a
number of materiel condition issues
arose during the 'inspection period
that required the plant personnel
to take prompt action and/or resulted
in TS
LCO entries.
The inspectors
considered
the licensee's
response
to
these materiel condition issues
to be appropriate.
While each
individual occurrence
was of minor consequence,
collectively the issues
represented
distractions for operators
and other plant staff.
In each
case,
the issue
was entered into the plant's maintenance
process
or
corrective action process
and corrected,
as appropriate.
The examples
are listed below:
From April 25,
1996, to Hay 14,
1996, the
"D" well water chlorination
system
was out of service for maintenance
due to system leaks
on one
train and
a
pump motor problem on the other train.
Without chlorination
of well water, drywell cooler fouling occurs
over time.
Between
April 25 and
Hay 14, drywell average
temperature
increased
from
approximately
118'F to 127'F.
Although this was below the
TS limit of
135'F, the inspectors
were concerned that chronic problems with the "D"
well water chlorination system continued to challenge
the plant.
A
similar concern with the well water chlorination system
was discussed
in
inspection report 50-331/96002.
~
On Hay 2,
1996, operators
received
a spurious
1/2 scram
on
Channel Al.
This was only the second
1/2 scram this year;
however,
the January
1996 1/2 scram also occurred
on Channel Al.
The licensee
was unable to determine
the cause,
and plans
were
being
made to connect
an event recorder to the channel for future
monitoring.
On May 16,
1996, the "B" reactor recirculation motor generator
(RRHG) set tripped.
This was the fifth trip of the unit in
2 years.
Operators
responded
well to the event
and successfully
maintained the plant in single-loop operations.
The licensee
was
unable to determine
the root cause of the trip.
Reference
Section
H1.4 for inspector
concerns
associated
with
troubleshooting of the event.
The inspectors will continue to
track and monitor this issue
under previously opened
IFI 50-
331/96002-01.
On May 23,
1996, while restoring the "B" RRHG set to service,
control
room operators
experienced
several
lock-ups of the "B"
RRHG set
scoop tubes.
Operators
were able to manually control the
unit and returned
the plant to full power.
Upon replacement
of
faulty deviation meters
in the circuitry, the
RRHG set functioned
properly.
~
On Hay 30,
1996, operators
declared
inoperable for the third time in the past
few months.
The
licensee initiated
an
AR to document evaluation
and resolution
and
was working with General
Electric to. understand
the cause.
c.
Conclusions
The inspectors
noted appropriate
operator
response
to the "B" RRHG set
trip and to the scoop tube lock-ups during restorations.
Engineering
support for short-term resolution of identified materiel condition items
was appropriate.
Engineering
and maintenance
activities for long-term
resolution of the
RRHG set trips and well water chlorination problems
were not effective in resolving the issues.
N8
Miscellaneous
Maintenance
Issues
(92902,
92700)
H8. 1
Closed
LER 50-331
96002
Revision 0:
Primary Containment Isolation
System
(PCIS) Half Group III Isolation
Due to Blown Fuse During
Maintenance.
The licensee
determined that the
PCIS isolation
was
a
normal
response
to the blown fuse in radiation monitor RM4116B.
The
fuse
was blown during
a maintenance activity in the
RH4104 cabinet
(located directly above
RH4116B)
when
a cable fell into the
RH4116B
cabinet.
Corrective actions,
which included inspection
and fuse
0
replacement
in RM4116B, reinforcing expectations
that loose cables shall
be secured,
and
a review for applicable training, were considered
appropriate.
This item is closed.
El
'a ~
Conduct of Engineering
Ins ection
Sco
e
37551
III En ineerin
El.l
Selected
engineering
problems or events
were evaluated to determine
their root cause(s).
,The effectiveness
of the licensee's
controls for
the identification, resolution,
and prevention of problems
was also
examined.
The inspection
included review of areas
such
as corrective
action systems,
root cause
analysis,
safety committees,
and self
assessment.
Verification of Mark I Hardened
Vent Modification
a ~
Ins ection
Sco
e
TI 2515
121
b.
The inspectors
used the guidance
in Temporary Instruction (TI) 2515/121
to review the licensee's
adherence
to commitments
made in response
to
Generic Letter (GL) 89-16, "Installation of Hardened
Wetwell Vent."
The
inspectors
reviewed
system diagrams,
Operating Instruction (OI),
description,
Emergency Operating
Procedures,
maintenance
history,
and
the Design
Change
Package
(DCP 1524),
and performed
a walkdown of
accessible
portions of the system.
Observations
and Findin
s
The inspectors verified that the as-built installation conformed to the
design criteria listed in TI 2515/121,
Appendix A, and that appropriate
plant procedures
and training were implemented.
The inspectors identified minor discrepancies
between the system valve
line-up in OI 573,
"Containment Atmosphere Control System," the system
diagram,
and valve labeling.
The discrepancies
were promptly corrected
when discussed
with the licensee.
The licensee
completed
a Quality Assurance
(QA) Surveillance of the
modification in September
1994
and identified one deficiency regarding
seismic qualification in the safety evaluation.
The inspectors verified
that the deficiency was promptly resolved.
The
QA Surveillance
was
considered
thorough
and detailed,
with appropriate
resources
assigned
to
the review.
10
c.
Conclusions
The inspectors
concluded that the licensee
met the commitments to
and that the hardened
vent was installed
as designed.
This
review closes
El.2
En ineerin
Su
ort to
RRMG Troubleshootin
Efforts
b.
Observations
and Findin
s
As documented
in Section Hl.4b, potentially valuable data
was lost
during troubleshooting efforts initiated in response
to the trip of the
"8"
RRHG set.
Engineering
support to data recovery
and trouble-shooting
efforts was weak.
c.
Conclusions
Engineering efforts to date
have
been ineffective with respect
to long-
term resolution of the
RRHG set trips.
E2
Engineering
Support of Facilities
and Equipment
E2. 1
S ent Fuel
Pool
Licensin
Basis
Review
a.
Ins ection
Sco
e
The staff identified in April 1996, that the Updated Final Safety
Analysis Report
(UFSAR) write up on spent fuel pool cooling did not
reflect the practices
at
DAEC as described
in a spent fuel pool rerack
submittal to the staff (Amendment
195 to the license).
There
appeared
to be areas of the
UFSAR description that should
be amplified to provide
design
bases
information to facilitate review of the current procedures
that assure
cooling for the spent fuel pool.
There also were areas
where the spent fuel pool rerack submittal
and the
UFSAR differed and
needed to be reconciled.
b.
Observations
and Findin
s
E2.1.1
In reviewing the licensing
bases of the spent fuel pool, the staff
determined that
DAEC had stated that under certain circumstances
it would use the
RHR system in the spent fuel pool cooling mode if
heat loads
were high enough to warrant it.
However,
DAEC did not
provide the equilibrium (i.e., bulk) temperature for when
RHR was
used to independently
provide cooling in lieu of the spent fuel
pool cooling
(SFPC)
system.
Based
on the rerack submittal,
the
temperature
should
be for the maximum heat load with one
RHR pump
and heat
exchanger
operating.
This inconsistency will be reviewed
further by the Office of Nuclear Reactor Regulation
(NRR)
as part
of the Spent
Fuel
Pool Licensing Basis
Review and is tracked
as
Inspection
Followup Item (IFI) 50-331/96004-04.
11
E2.1.3
E2.1.4
E2.1.5
In reviewing the licensing bases,
the staff noted that there are
a
number of values given for the
"maximum" temperature of the spent
fuel pool.
Section 5.4 (iii) of the rerack submittal states that
peak spent fuel pool temperature
is intended to be limited to
180
F assuming
two cooling trains,
which in turn is below the
212'F regulatory limit.
However in the
UFSAR, it appears
that the
SFP structural
design limit under normal
heat load conditions
(i.e.,
a temperature
to which it would be acceptable
to cycle
a
large
number of times) is analyzed to not exceed
150'F (the
temperature
found in the'merican
Concrete Institute Code).
In
addition, the calculated
maximum temperature for the full core
offload case
was 164.4'F,
which is above the 150'F in the
but below the
180
F limit.
These inconsistencies will be reviewed
further by
NRR as part of the
SFP Licensing Basis
Review and are
tracked
as IFI 50-331/96004-05.
In reviewing the licensing bases,
the staff found that the heat
load stated
by DAEC in the rerack submittal for the
Case 3'ull
core offload (18.87 x 10'tu/hr) is not consistent with
Supplement
1 of the rerack submittal
(18.73 x 10'tu/hr).
This
inconsistency will be reviewed further by
NRR as part of the
Licensing Basis
Review and is tracked
as IFI 50-331/96004-06.
The
UFSAR states
that
a fuel shuffle will not begin prior to
160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> after shutdown or that
a full core offload will not
begin prior to 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after shutdown.
However, the rerack
submittal states
that fuel would not be moved prior to 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />
after shutdown.
In addition, the Operating Instructions
(OIs) at
DAEC do not mention specific delay times that need to be honored
prior to moving fuel.
These inconsistencies will be reviewed
.
further by NRR as part of the
SFP Licensing Basis
Review and are
tracked
as IFI 05-331/96004-07.
The
DAEC spent fuel pool rerack submittal specifies
the maximum
number of fuel assemblies
moved out of the core,
put back in the
core, left in the spent fuel pool
(SFP),
and the number of new
assemblies
placed in the core during
a core reload.
However, the
applicable
procedures
do not specify maximum or minimum values of
,assemblies
to be moved.
This inconsistency will be reviewed
further
by NRR as part of the
SFP Licensing Basis
Review and is
tracked
as IFI 50-331/96004-08.
'ase
3: This scenario
corresponds
to the actual
discharge
practice
at
DAEC.
The calculations consider
a total of 3152 locations for the fuel storage
in the
pool
and
are carried
out at the point in time
when the stored
fuel
inventory is such that the addition of a normal batch to the pool will leave it
with insufficient capacity to accept
another
batch while maintaining the full
core discharge
reserve capability.
The transfer to the pool begins after
120
hours of in-core decay
and is conducted
at
144 assemblies
per
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Two
spent fuel pool cooling trains are
assumed
to be in operation.
12
E2.1.7
E2.1.8
In Section 9.1.3 of the
DAEC uses
the terms
"maximum normal
heat load" and
"maximum possible
heat load."
"Maximum normal heat
load" appears
to refer to fuel shuffling with approximately
1/3 of
the core being unloaded into the SFP.
The
"maximum possible
heat
load" appears
to refer to a full core offload.
Both terms
appear
to involve some conservative
assumptions,
but the
UFSAR contains
few details of the underlying analyses.
The assumptions
in the
terms
"maximum normal heat load"
and
"maximum possible
heat load"
do not correspond
to any of the four scenarios
detailed in the
.
1993 rerack submittal.
In OI 149
(RHR system),
pg.
54, the
definition of maximum normal
heat load disagrees
with the current
practice at
DAEC (i.e., full core offloads).
These
inconsistencies
will be reviewed further by NRR as part of the
Licensing Basis
Review and are tracked
as IFI 50-331/96004-09.
The
UFSAR defines the
RHR system in the Spent
Fuel
Pool Cooling
mode
as the alternate
cooling path for the spent fuel pool.
This
. path
has
a very complicated
and potentially time consuming
procedure.
Use of the
RHR system in the Spent
Fuel
Pool Cooling
mode has not been tested
since plant pre-operational
testing in
the early 1970s.
This inconsistency will be reviewed further by
NRR as part of the
SFP Licensing Basis
Review and is tracked
as
IFI 50-331/96004-10.
In Section 1.8. 13 of the
DAEC states
that
"The
RHR system
is used
as the source of makeup water [for the spent fuel pool]
and is classified
as Seismic Category I.
The Seismic Category I
piping is extended into the fuel pool cooling system
as far as
necessary
to ensure
the makeup water will get into the pool'."
This statement
is incorrect
and does not reflect other portions of
the
UFSAR (such
as Section 9.1).
No plant procedures
existed for
transferring water from RHR to the spent fuel pool.
These
inconsistencies
will be reviewed further by NRR as part of the
Licensing Basis
Review and are tracked as,IFI 50-331/96004-11.
E2.2
Standb
Li uid Control
Boron Solution 0 er atin
Conce tratio
b.
Observations
and Findi
s
On November
7,
1995, the licensee identified a discrepancy
between the
UFSAR and plant procedures for the standby liquid control
(SLC) system.
Section 9.3.4.2 of the
UFSAR specifies that maximum boron solution
operating concentration
is 14.6 weight percent,
however, this upper
concentration
limit was not in plant procedures.
For example,
monthly
surveillance
procedure
STP 44C001,
"SLC System
Boron Concentration
Test," Revision 3, which specified
a lower concentration limit, but not
an upper limit.
The inspectors
and licensee
reviewed past performance of STP 44C001
and
found only one instance
where the
UFSAR 14.6 weight percent requirement
was exceeded.
On January
10,
1992, test results
showed that the boron
solution concentration
was 14.93 weight percent;
however, there were no
13
E8
ES. 1
consequences
because
the temperature
at the time was
above the minimum
solution temperature
curve in TS.
The inspectors
were concerned that
there
were
no controls to prevent exceeding
the
maximum concentration
value because
the
SLC tank low temperature
alarm setpoint
was
based
on
a
maximum of 14.6 weight percent.
The licensee
revised the
and other
procedures
to specify the maximum concentration
value.
The inspectors
considered
the corrective actions to be appropriate.
This issue will be
reviewed further as
an IFI 50-331/96004-12.
Hiscellaneous
Engineering
Issues
Closed
Violation 50-331
94017-01:
Failure to Revise
Procedures
Following Changes
to Radiation Honitor Setpoints.
The cause
was
determined to be inadequate
controls in the engineered
maintenance
action
(EHA) process.
The inspectors
reviewed the licensee's
corrective
actions
and commitments
documented
in a letter to the
NRC dated
December
5,
1994.
The corrective actions
were considered
narrow in
scope
when
an additional
example of inadequate
controls in the
EHA
process
was identified on January
25,
1996.
As discussed,
in inspection
report 50-331/96002,
this was
a violation of 10 CFR Part 50, Appendix B,
Criterion XVI.
The inspectors will review the controls in the
EHA
process
and the implementation of corrective actions specified in the
licensee's
response letter to the
NRC dated
Hay 9,
1996, during the
closure of NOV 50-331/96002-07.
This item is closed.
IV Plant
Su
ort
Radiological
Protection
and Chemistry Controls
a.
Ins ection
Sco
e
In accordance
with procedure
71750,
selected activities associated
with
radiological controls, radiological effluents,
and waste treatment
were
reviewed to ensure
conformance with facility procedures
and regulatory
requirements.
Rl. 1
Increased
Trend in Off as Pretreat
Radiation
Release
Rate
b.
Observations
and Findin
s
Following restoration of the "B" RRHG set
and return of the plant to
full power, the inspectors
identified that the off-gas pretreat
radiation release
rate
had approximately doubled from earlier values
(from approximately 4-5 mrem/hour before the trip to approximately
8-10 mrem/hour after).
Control
room operators
were not aware of the
trend
and could not explain the reason for the increase.
The lead
chemist
had noticed the change
and
was investigating the issue,
but had
not communicated
his observations
to others.
The inspectors
were
concerned
that
a potential early indicator of a fuel problem was not
effectively communicated to others in the organization.
Subsequent
to
the exit meeting
on June
5,
1996, the licensee
demonstrated
to the
inspectors,
via coolant- sample results,
that the increase
in the level
was not attributable to
a fuel leak problem.
R1.2
Conclusions
in Radiolo ical Protection
and Chemistr
Control
The inspectors
had
no substantive
concerns with the licensee's
subsequent
efforts to resolve the increase
in offgas pretreat
release
rate,
as discussed
above,
once
a fuel leak was ruled out.
No other
concerns
were identified in this area.
V. Mana ement Neetin
s
Xl
Exit Heeting
Summary
The inspectors
presented
the inspection results to members of licensee
management
at the conclusion of the inspection
on June
5,
1996.
The licensee
acknowledged
the findings presented.
The inspectors
asked
the licensee
whether
any materials
examined during the
inspection
should
be considered
proprietary.
No proprietary information was
identified.
15
Licensee
PARTIAL LIST OF
PERSONS
CONTACTED
J.
Franz,
Vice President
Nuclear
G.
Van Hlddlesworth,
Plant Manager
R. Anderson,
Manager,
Outage
and Support
P. Bessette,
Manager,
Nuclear Licensing
J. Bjorseth,
Maintenance
Superintendent
J. Cantrell,
Manager,
Nuclear Training
D. Curtland, Operations
Supervisor
R. Hite, Manager,
Radiation Protection
M. McDermott, Manager,
Engineering
K. Peveler,
Manager,
Corporate guality Assurance
INSPECTION
PROCEDURES
USED
IP 37551:
Engineering
IP 61726:
Surveillance Observation
IP 62703:
Maintenance
Observation
IP 71707:
Plant Operations
IP 71750:
Plant Support
IP 92700:
Onsite Followup of Written Reports of Non-routine Events at Power
Reactor Facilities
IP 92901:
Followup Operations
IP 92902:
Followup Engineering
IP 92903:
Followup Maintenance
TI2515/121: Verification of Hark I Hardened
Vent Modifications
(GL 89-16)
ITEMS OPENED,
CLOSED,
AND DISCUSSED
~0ened
50-331/96004-01
50-331/96004-02
50-331/96004-03
50-331/96004-04
50-331/96004-05
50-331/96004-06
50-331/96004-07
50-331/96004-08
50-331/96004-09
50-331/96004-10
50-331/96004-11
50-331/96004-12
Closed
IFI
IFI
IFI
IFI
IFI
IFI
IFI
IFI
'FI
50-331/94003
LER
50-331/94006
LER
50-331/96002
LER
50-331/94017-01
16
LIST OF ACRONYHS USED
i
CFR
Code of Federal
Regulations
CHAR
Corrective maintenance
action request
Duane Arnold Energy Center
Design
change
package
EHA
Engineered
maintenance
action
End of cycle
GL
Generic Letter
IFI
Inspection followup item
IR
Inspection report
LCO
Limiting Condition for Operation
LER
Licensee
Event'Report
HD
Haintenance
Directive
Office of Nuclear Reactor Regulation
,
Operating Instruction
OSS
Operations Shift Supervisor
gA
guality Assurance
Residual
heat
removal
Reactor protection
system
Recirculation
pump trip
RRHG
Reactor recirculation motor generator
Spent fuel pool
SFPC
Spent fuel pool cooling
Surveillance Test Procedure
TI
Temporary Instruction
TS
Technical Specification
UFSAR Updated Final Safety Analysis Report
Unresolved
item
17