ML18036B255
| ML18036B255 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/14/1993 |
| From: | Kellogg P, Patterson C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18036B253 | List: |
| References | |
| 50-259-93-07, 50-259-93-7, 50-260-93-07, 50-260-93-7, 50-296-93-07, 50-296-93-7, NUDOCS 9304300055 | |
| Download: ML18036B255 (27) | |
See also: IR 05000259/1993007
Text
P0
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323
Report
NoseI
50-259/93-07,
50-260/93-07,
and 50-296/93-07
Licensee:
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 Units 1,
2,
and
3
Inspection at Browns Ferry Site near Decatur,
Inspection
Conducted:
February
18
March 19,
1993
Inspector:
a terson,
Sen>or
esi
ent
nspector
ate
igne
Accompanied
by:
J.
Hunday,
Resident
Inspector
R. Husser,
Resident
Inspector
J. Hathis, Project Inspector
E.
, License
Examiner
Approved by:
au
Re
tor Proj ct
, Section
4A
Division of Reactor Projects
SUMMARY
ate
Soigne
Scope:
This routine resident
inspection
included surveillance
observation,
maintenance
observation,
operational
safety
- verification, modifications, Unit 3 restart activities,
radiological controls,
and reportable
occurrences.
One hour of backshift coverage
was routinely worked during the
work week.
Deep backshift inspections
were conducted
on
February
21,
February
22,
and March 6,
1993.
9304300055
9304ib
" ADOCK 05000259
Unit 2 was in day 50 of a
100 day refueling outage at the end of
the report period,
paragraph
four.
The schedule
was
changed
from
119 days to 100 days
due to the shutdown .of other
TVA nuclear
plants.
No work scope reductions
were made.
One violation was identified for failure to comply with radiation
protection procedures,
paragraph
seven.
Three
examples of failing
to comply with radiation work permits were observed
by a
NRC
inspector during
a single day.
. One unresolved
item was identified concerning the need for a
10
CFR 50.59 evaluation for the recirculation
pump shaft replacement,
paragraph five.
The licensee
procedure
allows replacement
of a
like for like component without an evaluation.
However, the
procedure
may not be adequate
to address
the specifics
from a
safety standpoint.
This issue will require further evaluation to
determine
the adequacy of the licensee's
program.
REPORT DETAILS
Persons
Contacted
Licensee
Employees:
- J
- O
J.
- J
- R
D.
- M
- J
- M
+A.
- C
- Q
- J
A.
Bynum, Vice President,
Nuclear Operations
Zeringue,
Vice President
Scalice,
Plant Manager
Rupert,
Engineering
and Modifications Manager
Baron, Site guality and Licensing Manager
Nye, Recovery Manager
Herrell, Operations
Manager
Maddox,
Engineering
Manager
Bajestani,
Technical
Support Manager
Sorrell, Special
Programs
Manager
Crane,
Maintenance
Manager
Pierce,
Acting Licensing Manager
Corey, Site Radiological
Control Manager
Brittain, Site Security Manager
Other licensee
employees
or contractors
contacted
included licensed
.
reactor operators,
auxiliary operators,
craftsmen,
technicians,
and
public safety officers;
and quality assurance,
design,
and engineering
personnel.
NRC Personnel:
- P'. Kellogg, Section Chief
- C. Patterson,
Senior Resident
Inspector
- J. Munday, Resident
Inspector
- R. Musser,
Resident
Inspector
- J. Mathis, Project Inspector
"Attended exit interview
and initialisms used throughout this report are listed in the
last paragraph.
Surveillance Observation
(61726)
'
The inspectors
observed
and/or reviewed the performance of required SIs.
The inspections
included reviews of the SIs for technical
adequacy
and
conformance to TS, verification of test instrument calibration,
observations
of the conduct of testing,
confirmation of proper removal
from service
and return to service of systems,
and reviews of test data.
The inspectors
also verified that
LCOs were met, testing
was
accomplished
by qualified personnel,
and the SI's were completed within
the required frequency.
The following SIs were reviewed during this
reporting period:
O-SI-4.10.c.2,
Fuel
Pool Coolant Chemistry
On February
21,
1993, the inspector
performed
a review of
previously performed surveillance instructions for the chemical
analysis of the spent fuel pool.
This procedure,
O-SI-4.10.c.2;-
satisfies
TS requirement
4. 10.c.2 for all three units.
More
specifically, the fuel pool coolant is analyzed for chloride ion
concentration
and conductivity.
The limits -for conductivity and
chloride ion concentration
specified in the
TS are
10 pS/cm
and
0.5
ppm respectively.
While reviewing the data from the February
20,
1993 analysis,
the chloride ion concentration for all three
units was entered
as greater
than
.5ppm (approximately
.Sppm).
Previous
days readings
were noted to be much lower (by a factor of
one thousand).
The inspector
brought this matter to the attention
of the Chemistry Shift Supervisor.
The supervisor
informed the
inspector that the entries
in question
were in error in that the
actual chloride ion concentration
analyses
were performed in ppb
and the chemistry technician
had not converted the data to ppm.
The supervisor
on shift informed the inspector that the data would
be corrected to reflect the proper units.
The improper data
had
been
approved
as satisfactory
by the
technician
and the chemistry shift supervisor.
The inspector
, discussed
this matter with the Chemistry Superintendent
the
following morning.
The Chemistry Superintendent
indicated that
the data
had
one final review to be performed before being turned
into the surveillance coordinator.
He felt that this deficiency
would have
been detected
at this time.
To prevent another
occurrence of this type, the Chemistry Superintendent
discussed
this matter with the appropriate
chemistry personnel
and changed
procedure 0-SI-4. 10.c.2 to specifically convert the chloride ion
concentration
from ppb to ppm.
The inspectors will continue to
monitor surveillances
performed within the chemistry area.
SI-4.2.K.2.A(FT), Reactor Building Vent Exhaust Monitor
2-RH-90-250, Detector Channel
Functional Test
On March 11,
1993 the inspector
observed
the performance of
portions of the SI-4.2.K.2.A(FT), Reactor Building Vent Exhaust
Monitor 2-RH-90-250, Detector Channel
Functional Test.
This test
provides for the instrument functional test of the Reactor
Building Exhaust
Noble Gas Monitor Detector Channel
and partially
the requirements
specified in TS Tables 3.2.K and 4.2.K.
The
inspector noted that the current revision of the procedure
was
being used
and was being followed properly.
The test performers
appeared
knowledgeable
on the system
and the procedure.
The
inspector reviewed the completed surveillance
procedure
and noted
the surveillance
was completed satisfactorily
and
had received the
appropriate
reviews,
The inspector noted
no deficiencies
in this
area.
No violations or deviations
were identified in the Surveillance
Observation
area.
1
Haintenance
Observation
(62703)
Plant maintenance activities were observed
and/or reviewed for selected
safety-related
systems-
and components
to ascertain
that they were
conducted
in accordance
with requirements.
The following items were
considered
during these
reviews:
LCOs maintained,
use of approved
procedures,
functional testing and/or calibrations
were performed prior
to returning components
or systems
to service,
gC records
maintained,
activities accomplished
by qualified personnel,
use of properly
certified parts
and materials,
proper
use of clearance
procedures,
and
implementation of radiological controls
as required.
'I
Work documentation
(HR,
WR,
and
WO) were'eviewed
to determine
the
status of outstanding
jobs
and to assure
that priority was assigned
to
safety-related
equipment
maintenance
which might affect plant safety.
The inspectors
observed
the following maintenance activities dut ing this
reporting period:
a
~
Hydraulic Control Unit Haintenance
Op Harch 9,
1993 the inspector
observed
maintenance
being
performed
on hydraulic control units 42-51
and 46-51.
The scram
pilot valves were being rebuilt and their internals
replaced.
The
'aintenance
personnel
were knowledgeable
about the task
and
had
no
difficulties in its performance.
The job foreman
was present
to.
verify foreign material
exclusion
and proper placement of valve
internals.
The inspector noted
no discrepancies.
b.
Underground
Leak
During this period
an underground
leak developed
at the north side
of the
RHRSW building.'
tower crane
had
been
parked in this
location
and was
assumed
to be the cause of the leak.
The
inspector
was
aware that the 'tower crane loading
had
been
reviewed
as part of the lifting performed for the Unit 3
CCW pump
refurbishment.
The inspector
reviewed
DCN S-17560
and calculation
CD-(I0303-
921562,
4100M Crawler Hounted
Crane Evaluation for Lifting Unit 3
CCW Pumps.
The inspector
was particularly concerned
that
underground
safety piping as the
RHRSW piping could be the source
of water.
The
DCN determined that the lifting was acceptable
provided the crane
was located at a'ertain position indicated
on
a drawing
and the foundation
was prepared.
The preparation
consisted of placement
and compaction of crushed
stone and-
placement of timbers underneath
the crane tracks for load
distribution.
The foundation preparation
was necessary
because
of
the combined weight of the
CCW pump and crane.
V
'
However,
in this case
the unloaded
crane
was not parked to use the
benefit of the foundation preparation.
On page
25 of the
calculation,
the worst case
loading
on the
CCW conduit
w'as
evaluated
to occurr when the crane
was not loaded
and the matting
was not considered.
This would occur
as the crane
was
moved into
or out of position.
The loading of the crane,
in this worse
case
condition,
was still below loading of a proposed railroad tracks
considered
in the original design.
The licensee
moved the crane
away from the north side of the
building near the road.
An incident investigation
was initiated
for the leak.
, In the calculation the following embedded
items were reviewed:
1)
intake conduits
2)
18 inch
EECW pipe
3)
24 inch
RHRSW pipe
4)
3 inch demineralized
water pipe
5)
12 inch drain pipe
It was determined that both the
18 inch
EECW pipe
and
24 inch
RHRSW pipes
have protective sleeves.
The
18 inch
EECW pipe has
a
24 inch di'ameter sleeve.
The 24 inch
RHRSW pipe= has
a 30 inch
diameter'sleeve.
The'alculations
performed concluded that the
stresses
were significantly below the allowed.
The inspector
concluded that the crane loading
on the
embedded
piping and
particularly safety piping had
been considered.
The inspector
will continue to follow this issue with completion of the leak
repair
and II completion.
No violations or deviations
were identified in the Maintenance
Observation
area.
Operational
Safety Verification (71707)
The
NRC inspectors
followed the overal,l plant status
and any significant
safety matters related to plant operations.
Daily discussions
were held
with plant management
and various 'members of the plant operating staff.
The inspectors
made routine visits to the control
rooms.
Inspection
observations
included instrument readings,
setpoints
and recordings,
status of operating
systems,
status
and alignments of emergency
standby
systems,
verification of onsite
and offsite power supplies,
emergency
power sources
available for automatic operation,
the purpose of
temporary tags
on equipment controls
and switches,
alarm
status,
adherence
to procedures,
adherence
to LCOs, nuclear instruments
operability, temporary alterations
in effect, daily journals
and logs,
stack monitor recorder traces,
and control
room manning.
This
inspection activity also included
numerous
informal discussions
with
operators
and supervisors.
General
plant tours were conducted.
Portions of the turbine buildings,
each reactor building',
and general
plant areas
were visited.
Observations
included valve position
and system alignment,
and
hanger conditions,
containment isolation alignments,
instrument
readings,
housekeeping,
power supply
and breaker alignments,
radiation
and contaminated
area controls,
tag controls
on equipment,
work
activities in progress,
and radiological protection controls.
Informal
discussions
were held with selected
plant personnel
in their functional
areas
during these tours.
a
~
Unit Status
Unit 2 was in day 50 of a
100 day refueling outage at the
end of
the report period.
The outage
schedule
was changed
from 119 days
to 100 days during this period.
The next major milestone will be
e'stablishment
of secondary
containment
scheduled for April 10,
1993.
b.
C.
Clearance
Tag Placement
on Electrical
Breakers
While touring the turbine building on March 3,
1993,
the inspector
noted
an electrician working inside
a 480 volt breaker
which had
a
clearance
tag hanging
on the compartment
door.
The inspector
had
been told by the Operations
Manager.
several
weeks before that if a
breaker door had
a clearance
tag hanging
on it, the door could not
be opened.
The inspector questioned
the electrician
and
he stated
that this was true for 4
kV breakers
but not for 480 volt
breakers.
A memorandum written by the Operations
Superintendent
dated
February
7,
1993, stated that if a clearance
tag is placed
on
a breaker
compartment
door, the door becomes
the clearance
boundary.
The Maintenance
Manager
was also
asked
about accessing
a breaker
compartment with a clearance
tag hanging
on it and
he
stated that it was all right.
He further stated that discussions
held with the Operations
Superintendent
resulted
in the
aforementioned
memorandum
and the intent was to allow maintenance
to access
breakers
with clearance
tags
hanging
on the door so that
maintenance
could
be performed.
SSP-12.3,
Equipment Clearance
Procedure,
does not contain instructions that specify where to
hang the tag
on breakers.
This matter
was brought to the
attention of the Operations
Manager
who generated
a new memorandum
dated
March 15,
1993,
which stated that if a breaker
door= had
a
clearance
tag
on it, the door could not be opened.
If access
to
the inside of the compartment
was
needed
Operations
would move the
tag to the inside.
This memorandum
should provide the additional
guidance
needed to ensure
safe operation
and breaker
maintenance.
Annual Operating
Report
The inspector
reviewed the Browns Ferry Nuclear Plant Units 1, 2,
and
3 Annual Operating
Report for January
1,
1992 through
December
31,
1992.
The report included
a summary of safety evaluations for
FSAR changes,
procedure
changes,
special
operating conditions,
6
E
special
tests,
temporary alterations,
and plant modifications.
It
also included the
1992 Radiological
Release
Summary,
Occupational
Exposure
Data,
Challenges
to Hain Steam Relief Valves,
and the
Reactor
Vessel
Fatigue
Usage Evaluation.
This report satisfies
the
requirements
of 10
CFR 50.59,
Regulatory
Guide 1. 16 Sections
1.b.(1),
(2),
and
(3)
and
TS Sections 6.9. 1.2
and 6.9.2. 1.
Housekeeping
During the inspection period,
the inspector
performed
an audit of
the licensee's
control of housekeeping.
As a part of this audit,
the inspector evaluated
the'implementation
of procedure
SSP-12.7,
Housekeeping/Temporary
Equipment Control.
This site standard
practice
procedure
delineates
the housekeeping
control practices
and requirements for the plant.
One of the specific requirements
of the procedure
is the designation of a material
control/housekeeping
coordinator.
As
a part of the inspection
effort, the inspector
reviewed the housekeeping
program with this
individual.
The housekeeping
coordinator
and the inspector discussed
the
plant's
housekeeping
program in detail.'
major aspect of the
housekeeping
program is the inspections
performed
by plant
personnel.
The Unit 2 reactor building, turbine building, diesel
generator
buildings
and the intake structure
are divided into 35
zones for daily inspection.
Each
zone
has
a zone inspector,
whose
name is prominently displayed in the zone,
tasked with this daily
inspection.
Once
a week, the inspection of the various
zones
are
documented
on Appendix
C of SSP 12.7
and turned into the
housekeeping
coordinator.
This documentation
is to include
deficiencies
discovered
and associated
corrective actions.
The
appendix
(C) contains
a comprehensive list of housekeeping
deficiencies that the zone inspectors
are to use
as
a guide
when
inspecting their zones.
The inspector
reviewed the results of these
inspections for the
period of November ll, 1992 - Harch 8,
1993.
A mixture of
inspection results
was noted.
While it appeared
that
many zone
inspections
are performed
and documented
thoroughly, other
housekeeping
inspection reports .indicated that documentation
of
inspections
was weak.
The inspector
expressed
this concern to the
housekeeping
coordinator.
The coordinator stated
her awareness
of
this matter
and indicated that the zones
which appeared
from
review of inspection
documentation
to get the least attention
during inspections
were frequently chosen to be inspected
during
the plant managers
weekly walkdown.
Another'spect of this evaluation
was th'e inspector's
walkdown of
the plant for general
housekeeping
practices.
Currently, Unit 2
is in the midst of its cycle
6 refueling outage
and
a great
amount
of equipment is spread
throughout the plant.
Host prevalent of
this equipment is scaffolding.
The inspectors will continue to
tour the plant during the outage
and ensure that the majority of
scaffolding is removed
from above safety related
equipment prior
to startup.
Other items noted during plant tours were welding
bottles that are not currently being utilized.
These
items were
brought to the attention of the licensee.
The overall condition
of housekeeping
for the last 2-3 months
appears
to be in somewhat
of a decline
due to outage activities in Units
2 and 3.
The
inspectors will continue to inspect the licensees'ousekeeping
program to ensure plant conditions
are brought
up to pre-outage
standards.
As the current outage
comes to an end,
the inspectors
will more closely monitor housekeeping
practices
and conditions in
order to help ensure
proper
operation of plant equipment.
e.
Fire Door Blocked Open
On March 10,
1993, at approximately
1730, during
a routine tour of
the control
bay, the inspector
noted that door ¹464,
the Computer
Room fire door on the
1C elevation,
was cracked
open.
The
internal door knob was missing
and the door blocked. open
so that
personnel
could exit the computer
room.
Since the door was fire
rated,
the inspector
asked the
ASOS whether
a
LCO had
been written
on the door.
The
ASOS requested
plant fire protection
personnel
to make this determination.
Plant fire protection determined that
a
LCO was not in effect for the door in question.
Fire Protection
personnel
initiated plant form "Attachment
F" documenting
the
condition.
as required
by TS 3. 11.G. l.a,
was established.
At the time of the event, all three
Browns Ferry
units were defueled
and this event
was of minor safety
significance.
Plant personnel
need to be reminded that fire rated
assemblies
should not be defeated
without the proper compensatory
actions taken.
Spent
Fuel
Pool
The inspector performed
a review of the licensee's
controls for
the spent fuel pool during the inspection period.
The inspection
effort was performed to ensure that adequate
controls were in
place for the control of spent fuel pool parameters
and that
TS
requirements
for the
SFP being met.
A complete core off load had
been
completed at the
end of the previous inspection period for
the Unit 2 cycle
6 refueling outage.
The inspector
reviewed
TS surveillance
requirements
for the spent
fuel pool water.
TS 4. 10.c. I requires that whenever irradiated
fuel is stored in the spent fuel pool, the water level
and
temperature
shall
be recorded daily.
The inspector verified that
the temperature
of the pool
was being recorded daily in accordance
'ith
procedure
2-SI-2,
Instrument
Check
and Observations.
The
same procedure
contained
the requirements for the recording of the
SFP water level,
however, it did not specify that
a specific water
level
be recorded.
Rather than record
a specific water level, the
procedure directs operations
personnel
to check
a control
room
8
("Fuel
Pool
System Abnormal" ) and if the annunciator
is not illuminated, record the
SFP water level 'as normal.
The
fuel pool level switches
and skimmer surge tank level switches
input into this annunciator
and
no direct reading of fuel pool
level is currently available.
The
NRC
questioned
the adequacy of
the
SFP level documentation
with respect
to TS 4. 10.c. l.
The
licensee
has
agreed
to submit
a
TS change to clarify the method
by
which SFP water level will be monitored.
No violations or deviations
were identified in the Operational
Safety
Verification area.
Modifications (37700,
37828)
The inspectors
maintained
cognizance of modification activities to
support the restart of Unit 2.
This included reviews of scheduling
and
work control, routine meetings,
and observations
of field activities.
Throughout the observation of modifications being performed in the field
gC inspectors
were observed
monitoring and documented verification at
work activities.
a
~
Fire Protection Modifications
The inspector
reviewed the
DCNs associated
with the fire protec-
tion system
upgrades.
This is
a commitment
made
by the licensee
to meet
NFPA standards.
The upgrade
includes installation of new
fire detection
systems
in the plant.
The
DCNs and location for
each
are
as follows.
W17911
WI7909
W17910
W17907
W17906
W17908
W17904
Local System at the Intake Structure
Local System in Unit I/2
DG Building
Local System in Unit 3
DG Building
Local System in Unit 2 Reactor Building
Local System in Cable Spreading
Room
Local System, in Unit 3 Reactor Building
Installs central
computers
and interconnection
of all local systems.
In addition,
W18213 will be implemented to power down
and
decommission
the existing detection
systems.
Typical of components
installed at each location are addressable
smoke detectors,
thermal detectors,
manual pull stations,
local
fire alarm and control panels,
and local horn/strobe
alarms.
In
the
DG building, additional
equipment
was provided for the carbon
dioxide systems.
After completion of the installation, there is
a testing window in
the schedule.
Each local panel will have
a loop checkout
performed
as
soon
as the local detection
systems
are functional.
After all work has
been completed,
there is
a four-day vendor
setup of the system followed by a
25 day post modification test
window.
The inspector
reviewed several
of the
DCN packages
with emphasis
on W17904 for the interconnection of all local systems.
These
activities will continue to be monitored
as the modifications are
worked.
Hardened
Wetwell Vent
Installation of the hardened
wetwell vent continued throughout the
inspection period.
The majority of the work effort in the Unit 2
reactor building involved the completion of the saddle
weld which
joins the
14 inch hardened
wetwell vent line to the existing
20
inch line from. the torus.
The licensee
experienced
numerous
problems in process
of completing the weld.
A number of failures
during
the licensee
to excavate
portions of
the joint for repair.
,Final
acceptance
of this weld joint
occurred
on March 18,
1993.
In addition, discrepancies
with the
licensee's
welding process
were identified during
a Region II
based
inspection
(see
IR 259,
260,
296/93-05)
performed during
this inspection period.
Work continued
on the outside
common portions of the-hardened
vent.
Excavation in the vicinity of the plant stack for a vent
drain line and its associated
valve pit progressed
during the
inspection period.
The inspectors will continue to monitor the
licensee's
work on the hardened
wetwell vent
and ensure
adequate
post modification testing is performed prior to the completion of
the outage.
Small
Line Cracks
During Unit 2 Cycle
6 operation three
small lines inside the
containment
experienced
cracks.
The inspector reviewed the
licensee
plans to mitigate the risk of similar events
in the
future.
Technical
support identified 43 test,
vent, drain,
and
instrument lines attached
to the recirculation
and
RHR lines
inside the containment.
Eight lines are
no longer needed.
These
will be cut and capped'he
remaining lines will undergo
an
analyses
of the existing configuration for adequate
supports.
Also,
a visual
and liquid penetrant
inspections of -the welds will
be performed to determine
any other necessary
repairs or
corrective action.
The inspector will continue to follow these
activities
as the systems
are returned to service.
Unit Battery
3 Replacement
During the inspection period; the inspectors
monitored the
replacement
of 250
VDC Unit Battery 3 in accordance
with DCN
W17274.
Unit Battery
3 is
a 120 cell
250
VDC which is being
replaced
in anticipation of multi-unit operation
and
because
the
10
existing battery cells were nearing the end of design life.
The
new battery cells have
a higher capacity than
do the existing
cells due to an increase
in the number plates
per cell.
Although,
the
new cells will contain
an increased
number of plates,
the size
of cell container wi.ll be the
same
as the existing cells.
The
increase
in number of plates
caused
a corresponding
increase
in
cell weight
and necessitated
the replacement
of the battery racks.
The inspector
reviewed the installation of the
new battery racks
and installation of the
new battery cells.
Testing of the
new
battery is expected
to occur in the near future
and will be
monitored
by the inspectors.
Recirculation
Pump Rotating Element
Replacement
The inspector
reviewed
and witnessed
the replacement
of the
recirculation
pumps rotating elements
during refueling outage
cycle
6 for Unit 2.
The shafts
were replaced
due to an industry
problem with thermal fatigue cracking.
The replacement
of .the
pump shafts
were done under work orders
92-6679300
(2B pump)
and
92-6640400
(2A pump) respectively.
The recirculation
pump design
had
been
upgraded
from the original
design.
The upgrade affected the rotating element,
cover/heat
exchanger,
hydrostatic bearing
and material
composition.
The
upgrades
were due to thermal cracking problems
experienced
by
other
BWRs and
PWRs.
During the review process
of the work packages,
the inspector
noted that
a safety evaluation
had not been
performed
by the
licensee
in accordance
with 10
CFR 50.59 requirements.
10 CFR Part 50, Appendix B, Criterion IV, states
that design
changes,
including field changes,
shall
be subject to design control
measures
commensurate
with those applied to the original design
and
be approved
by the organization that performed the original
design
unless
the applicant designates
another responsible
organization.
Furthermore,
measures
shall also
be established
for
the selection
and review for suitability of application of
materials,
parts,
equipment,
and processes
that are essential
to
the safety-related
functions of the structures,
systems
and
components.
The licensee
'considered
the upgrade of the
recirculation
pumps
a like-for-like replacement
thereby
a safety
analysis
was not warranted.
The equivalency test (fit, form and
function) is satisfied
according to the licensee.
Procedure
PI-89-06,
Design
Change Control, step 13.5.d states
that
if a replacement
item meets
the criteria where the item is part of
equipment
designed
by a vendor for the specific plant application
and the vendor certifies that the replacement
item is equivalent,
the item can
be ordered with no further justification.
The staff
agrees
that the upgraded recirculation
pump rotating element,
cover/heat
exchanger
and hydrostatic bearing
may meet the original
11
fit and function requirement of the equivalency test,
however it
does
not meet the form requirements.
The material for the shaft,
impeller, hydrostatic bearing etc... all have
been
changed.
The
physical materials
have
changed therefore
form requirements
are
not satisfied.
This issue requires further
NRC review and will be
tracked
as
URI 259,
260,
296/93-07-01,
Safety Evaluation for
Recirculation
Pump Shaft Replacement.
Unit 3 Restart Activities
(30702)
The inspector
reviewed
and observed
the licensee's
activities involved
with the Unit 3 restart.
This included reviews of procedures,
post-job
activities,
and completed field work; observation of pre-job field work,
in-progress field work, and gA/gC activities; attendance
at restart
craft level, progress
meetings,
program meetings,
and management
meetings;
and periodic discussions
with both
TVA and contractor
personnel,
skilled craftsmen,
supervisors,
managers
and executives,
a.
Unit Status
Limited activities continue
on Unit 3 recovery.
A schedule
review
process
is ongoing to determine
a credible schedule
to be
announced
at the end-.of the Unit 2 Cycle 6.outage.
Activities
were drywell steel
work and return to service of the
RWCU system.
b.
Drywell Tour
On March 5,
1993, the inspector
made
a tour of the Unit 3 drywell.
Overall the drywell was clean
and free of combustible material.
The inspector
noted that many hot jobs were in progress
which re-
quired firewatches.
Each job had its
own firewatch.
Blankets
and
catch
pans
were
used in many places to prevent slag from dropping
to
a lower elevation.
The inspector
reviewed the welding and
grinding permits posted
and verified the information required
was
documented
properly.
The inspector
found no deficiencies.
Radiological Controls
(83724)
a
~
Drywell Cameras
The Radiation Control
Group has installed approximately twenty
cameras
throughout the Unit 2 drywell which input to fifteen
monitors located at
a manned post outside the drywell.
In
addition,
intercom stations
have
been established
in the drywell
which can communicate with the manned post.
If a person
in the
drywell needs
assistance
they can talk via an intercom with the
person monitoring this post.
The monitor can then adjust the
camera to see the person requesting
assistance
and also alert the
HP stationed
in the drywell.
On 3/16/93,
the inspector toured the
drywell and noted that while this arrangement
is
a good idea,
the
stations
in the drywell are not easily identifiable.
In addition
the drywell radiological control technician
was not aware that
~
12
they even existed.
The cameras
also provide for a much larger
surveillance
area
by radiological controls without expending
any
additional radiation dose.
Three
VCRs are available which are
used to filivarious jobs or job sites
and then viewed outside the
drywell in low radiation areas
to resolve
problems that
may arise.
The films are also often used during shift change
or for training
purposes.
In addition to the drywell, cameras
have
been installed
in the steam tunnel
and the
RWCU heat exchanger
and
pump rooms.
The estimated
dose savings,
by the licensee,
resulting from the
use of the cameras
is 16.8 Nan-Rem.
Radiological
Control
Work Practices-
On February
25,
1993, during the performance of daily rounds,
the
inspectors
observed
work activities
on the 664'levation
(refuel,
floor) of the Unit 2 reactor building.
A particular work activity
observed
consisted of two individuals performing maintenance
on
the fuel support piece lifting tool.
While,one worker with a face
shield
was manipulating the hose
connected
to the lifting tool,
the other worker without
a face shield
was bending
down on his
knees
and handling the lifting tool.
Shortly after the
observation of this incident, the two workers switched their work
positions.
After holding
a discussion
with the radiological
control technician
on shift, the inspector identified that the
worker without the face shield was not signed
on the
RWP to
perform work activities
on the fuel support piece lifting tool.
Furthermore,
a face shield
was required per radiological control
directions to perform work activities
on fuel support piece
lifting tool.
Later the
same day, another inspector performing
a routine tour of
the Unit 2 turbine building observed
maintenance
being performed
on the turbine stop valves.
Individuals performing the work
activity were dressed
in anti-contamination clothing as specified
by radiological controls
and
RWP 93-2-60002-01-00
as the area in
question
was being controlled
as
a contamination
zone.
As the
maintenance
progressed,
the inspector
observed
(as did the
radiological control technician monitoring the job)
an individual
remove his anti-C hood
and surgeons
cap while still in the
contamination
zone.
This action
was taken prior to the individual
climbing from the "valve pit" to the C-zone exit.
The
radiological control technician
ensured that the individual exited
the C-zone,
undressed
and proceeded
to the frisking station.
A
few minutes later, the inspector
observed
another individual in
the C-zone
don
an anti-C hood which had
been lying on
a steam line
within the C-zone.
The inspector
informed radiological control of
his observation.
Radiological control instructed the individual
to exit the C-zone
and perform
a whole body frisk.
In both
instances,
the individuals were found not to be contaminated.
TS 6.8. I.l.a requires that written procedures
shall
be
established,
implemented
and maintained covering the applicable
13
procedures
recommended
in Appendix A of Regulatory
Guide 1.33,
Revision 2, February
1978.
Regulatory
Guide 1.33,,section
7.e. l.,
requires
Radiation Protection
Procedures
covering Access Control
to Radiation Areas including
a Radiation
Work Permit Systems.
RCI-9, Radiation
Work Permits,
is the implementing procedure for
this requirement.
Failure to comply with RCI-9, Radiation
Work
Permits,
section 6.5. 1, which holds the individual worker
responsible
to ensure
the correct
RWP for the job is used,
and
section 6.5.3.'which requires individuals using
a
RWP comply with
all of the requirements
of the
RWP as 'well as the verbal
instructions given by radiological control personnel
so far as
.
those instructions pertain to radiological matters,
is
a violation
of TS Section 6.8. 1, Procedures.
This matter is identified as
violation 259,
260, 296/93-07-02,
Failure to Comply with Radiation
Protection
Procedures.
One violation was identified in the radiological control work practices-
area;
8.
Reportable
Occurrences
(92700)
The
LER listed below was reviewed to determine if the information
provided met
NRC requirements.
The determinations
included the
verification of compliance with TS and regulatory requirements,
and
addressed
the adequacy of the event'escription,
the corrective actions
taken,
the existence of'otential generic
problems,
compliance with
reporting requirements,
and the relative safety significance of each
event.
Additional in-plant reviews
and discussions
with plant
personnel,
as appropriate,
were conducted.
(CLOSED)
Inadvertent
Spillage
This item was originally identified when in May 1986,
an inadvertent
- actuation of an
ESF occurred
in Unit I and
was twice repeated.
The
actuation
was caused
by a false high drywell signal
due to an electrical
short.
All eight
DGs and two
EECW pumps started automatically.
Since
CS and
RHR pump motor breakers
were tagged,
no
pumps started.
However, the
CS injection valves opened,
which allowed water from the
condensate
storage
system to flood the reactor cavity.
Water over
flowed into the vents
on the periphery of the refuel,ing well and
some
spillage occurred
from the ventilation ductwork on the lower elevations
of Unit I reactor building before the valves
were discovered
open.
The
electrical short,
caused
by moisture in two high drywell pr essure
switches,
was believed to be due to
a spurious actuation of fire spray
valves in the area of these
switches earlier in the week of the event.
Inspection
Report 90-27
and
addressed
this issue
and
closed this item for Unit 2 only.
The Unit 2 Reactor Building Fire
Spray System
had
been modified to a preaction type system which operates
on the fused
head
spray valve design.
When
an actuation
occurs,
the
system floods with water,
and only those
spray valves
where the fuse
head
has disengaged will actually spray water.
Spurious actuations will
only cause
the system to flood with water without actual
spraying.
The
systems for Unit I and Unit 3 had not been modified.
The inspector
was
concerned
whether
any equipment of Unit I and/or Unit 3 could affect
Unit 2 system operability.
Inspection
Report 92-16,
VIO 259,
260,
296/87-33-01,
Failure to Seal
Conduit,
addressed
that the licensee
took
actions to correct this problem with Appendix
R modifications
and the
sealing of all required conduits
and conjunction boxes.
Based
on the review of the closure
package,
applicable
LERs and
violation to this item, the inspector considers this
LER for Unit I and
Unit 3 closed.
Exit Interview (30703)
The inspection
scope
and findings were summarized
on March 19,
1993,
with those
persons
indicated in paragraph
I above.
The inspectors
described
the areas
inspected
and discussed
in detail the inspection
findings listed below.
The licensee
did not identify as proprietary
any
of the material
provided to or reviewed
by the inspectors
during this
inspection.
Dissenting
comments
were not received
from the licensee.
Item Number
Des'cri tion and Reference
259,
260,
296/93-07-01
259,
260,
296/93-07-02
URI, Safety Evaluation for Recirculation
Pump Shaft Replacement,
paragraph five.
VIO, Failure to Comply with Radiation
Protection
Procedures,
paragraph
seven.-
and Initialisms
DCNs
LER
LCO
PPB
Condenser Circulating Water
Chemistry Shift Supervisor
'irculating
Water
Design
Change Notices
Diesel Generator
Emer'gency
Core Cooling Systems
Emergency
Equipment Cooling Mater
Engineered
Safety Feature
Final Safety Analysis Report
In-Vessel
Visual Inspection
Licensee
Event Report
Limiting Condition for Operation
Maintenance
Request
National Fire Protection Association
Nuclear Reactor Regulation
Parts
Per Billion
Licensee
management
was informed that I
LER was closed.
PPH
TS
Parts
Per Million
Quality Assurance
Quality Control
Radiological
Control Instruction
Residual
Heat
Removal
Residual
Heat
Removal Service
Water
Reactor
Water Cleanup
Radiological
Work Permit
Spent
Fuel
Pool
Surveillance Instruction
Technical Specification
Unresolved
Item
.Violation
Work Order
Work Request
~,
~
0