ML18033B296
| ML18033B296 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/03/1990 |
| From: | Carpenter D, Little W, Patterson C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18033B294 | List: |
| References | |
| 50-259-90-08, 50-259-90-8, 50-260-90-08, 50-260-90-8, 50-296-90-08, 50-296-90-8, NUDOCS 9005170005 | |
| Download: ML18033B296 (24) | |
See also: IR 05000259/1990008
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323
Report Nos.:
50-259/90-08,
50-260/90-08,
and 50-296/90-08
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
Cond cted:
March
16 - April 16,
1990
Inspectors:
Accompanied
by:
Approved by:
.
C
p
ter
NRC Sit
nager
. Patterson,
NR
estart
oor snator
E. Christnot,
Resident
Inspector
W. Bearden,
Resident
Inspector
K. Ivey, Resident
Inspector
R. Bernhard
Project Engineer
p~
Date
igned
Ss&
Date
igned
it
e,
ection
C ief,
Inspection
Programs,
TVA Projects Division
ate Signed
SUMMARY
Scope:
This
routine
resident
inspection
included
surveillance
observations,
maintenance
observations,
operational
safety
verifications,
reportable
occurrences,
action
on previous
inspection findings, implementation of nuclear
quality assurance
plan,
high potential testing of electrical
cables,
and site
management
and organization.
0
90051700. 5
~00025
c)
900-04
pDF;
lg
Results:
One violation was identified for failure to take
a chemistry sample,
paragraph
4.
Although
an
employee
brought this error to the attention of management
there
have
been
several
missed
chemistry samples
during the past
two years.
A
NCV was identified for failure to adequately
control test measures
during high
potential
testing of electrical
cables,
paragraph
8.
The licensee
took prompt
corrective action to resolve this issue.
The closeout of tw'o unresolved
items
resulted
in two non-cited violations related to
HVAC design
and drywell beams
design,
paragraphs
6.g.,
and 6.h.
~
REPORT
DETAILS
Persons
Contacted
Licensee
Employees:
0. Zeringue, Site Director
L. Myers, Plant Manager
- M. Herrell, Plant Operations
Manager
- R. Smith, Project Engineer
J. Hutton, Operations
Superintendent
A. Sorrell, Maintenance
Superintendent
G. Turner, Site guality Assurance
Manager
P. Carier, Site Licensing Manager
- P. Salas,
Compliance Supervisor
J. Corey, Site Radiological Control Superintendent
R. Tuttle, 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 Employees
- D. Carpenter,
Site Manager
C. Patterson,
Restart Coordinator
- E. Christnot, Resident
Inspector
- W. Bearden,
Resident
Inspector
K. Ivey, Resident
Inspector
- Attended exit interview
Acronyms used throughout this report are listed in the last paragraph.
Surveillance
Observation
(61726)
The inspectors
observed
and/or
reviewed the, performance
of surveillance
testing
during this reporting period.
The inspections
consisted
of a
review of the
for technical
adequacy
and
conformance
to
TS,
verification of test instrument calibration, observation of the conduct of
the test,
confirmation of proper
removal
from service
and return to
service of the system,
and
a review of the test data.
The inspector also
verified that, limiting conditions for operation
were met, testing
was
accomplished
by qualified personnel,
and the
were completed at the
required frequency.
The following SIs were observed/reviewed:
2-SI-2,
Instrument
Checks
and Observations.
This SI ensures
that
instrument checks
and observations
required
by the
TS to be performed
on
a once
per shift, daily, or semi-weekly frequency
are
performed.
This SI is performed
by the operations staff.
The inspector
noted
that
there
were
only
a
few checks/observations
required
for
the
current plant conditions.
No deficiencies
were identified.
2-SI-3.7.A-1(A), Suppression
Chamber
Narrow Range
Level Instrumenta-
tion Channel
A Calibration.
This SI
was also validated
as it was
performed.
This included performing all steps
in the procedure
even
if the procedure itself did not require all steps
to be performed.
This method
ensures
that all of the procedure
steps
are validated
by
performance.
The
was
performed
as written
and
no procedural
changes
were
required for the validation.
No deficienci'es
were
identified.
No violations
or deviations
were
identified
in
the
Surveillance
Observation
area.
3.
Maintenance
Observation
(62703)
Plant
maintenance
activities
on
selected
safety-related
systems
and
components
were observed/reviewed
to ascertain
that they were conducted
in
accordance
with requirements.
The following items were considered
during
this review:
the limiting conditions for operations
were met; activities
were
accomplished
using
approved
procedures;
functional testing
and/or
calibrations
were
performed prior to returning
components
or system
to
service;
quality control
records
were
maintained;
activities
were
accomplished
by qualified
personnel;
parts
and
materials
used
were
properly certified; proper tagout clearance
procedures
were
adhered
to;
Technical
Specification
were
met;
and
radiological
controls
were
implemented
as required.
Maintenance
requests
were reviewed to determine
the status
of outstanding
work activities
and to assure
that priority was
assigned
to equipment
maintenance
which could affect plant safety.
The
inspectors
observed
the following maintenance
activities
during this
report period:
Preventive
maintenance
conducted
on the potential transformer for the
3ED shutdown
board normal feeder breaker.
High Potential Testing of Cables
At the beginning of this 'reporting period, the licensee
implemented
a
new
system for the conduct of maintenance.
This system is detailed
in
SDSP
7.6,
Maintenance
Management
System.
The
new system .includes
changes
to
the methods for requesting
maintenance
from the old paperwork
MRs to
a
card.
Valid
are
processed
into
WOs which provide the
necessary
information
and instructions
to perform the work.
The
new
WR cards
are
individually numbered
and
have
two sections
which can
be detached.
One
section
is
a large, bright orange
tag which is affixed to the equipment
requiring work.
The tag also includes
two small detachable
stickers that
can
be
placed
on small
devices
or equipment.
The other section
is
a
detailed
request
for work that
includes
the
component
identifier;
location;
problem
description;
method of discovery;
Operability,
LCO
0
0'
e
entry,
and
CARR evaluation
signoffs;
and work planning sections.
The
inspectors
noted several
of the
new tags
and stickers
on equipment in the
plant.
The
inspectors
considered
this
new
program
an
example of the
attention that the
new maintenance
organization is giving to establishing
good
work practices,
reducing
personnel
errors,
and eliminating
the
various
maintenance
problems
which have
occurred at Browns Ferry in the
past.
No violations or deviations
were identified in the Maintenance
Observation
area.
4.
Operational
Safety Verification (71707)
The
NRC inspectors
were kept informed of the overall 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.
0
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;
onsite
and offsite
emergency
power
sources
available
for
automatic
operation;
purpose of temporary
tags
on equipment controls
and
switches;
alarm status;
adherence
to procedures;
adherence
to
limiting conditions
for operations;
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
positions
and
system
alignment;
and
hanger
conditions;
containment
isolation
alignments;
instrument
readings;
housekeeping;
proper
power supply and breaker
alignments;
radiation area
controls;
tag controls
on equipment;
work activities
in progress;
and
radiation
protection
controls.
Informal discussions
were
held with
selected
plant personnel
in their functional areas
during these tours.
'a ~
Frequency
Reduction
Program
The licensee
has
established
a
SFRP.
The inspector
reviewed
the
status
of this program.
The program is defined in SDSP 12.10,
Frequency
Reduction
Program.
A long term goal for BFNP
has
been
established
of one unplanned
scram or less
per unit per reactor year.
The instruction outlines
the function, affected organizations,
and
goals of the
Frequency
Reduction
Team
and the
mechanism
the
SFRP will use to pursue
the scram reduction goal.
The first action
taken
by the
team
was to evaluate,
review,
and
determine corrective actions of all
BFNP scrams
from January
1978, to
March 1985,
on
an individual system
by system basis.
Another action
0
was
to evaluate
industry
recommendations
primarily from the
Owners
Group for applicability to
BFNP.
From these
reviews
and
evaluations,
122
recommendations
were
prepared
by the
SFRC
and
approved.
The status of the recommendations
is as follows:
7 of the
36
hardware
related
recommendations
are
closed,
and
42 of the
86
administrative related
ones
are closed.
One of the
improvements
made
was to develop
a procedure
to manually
verify continuity of MSIV solenoids prior to testing.
Some of the
recommendations
requiring modification are planned for next cycle.
A
modification
to install continuity indicators will be
performed
during the next refueling outage.
Although the benefits
and
success
of the
can only be
proven
during plant operations,
the
program
is
a positive step
by the
licensee.
b.
Loss of Thermal
Margin Caused
by Channel
Box Bow.
This information notice
was
intended
to alert
BWR sites of potential
problems
involving loss of thermal
margin caused
by excessive
bowing
of fuel
channel
boxes.
This channel
bowing resulted
in a modeling
error in the plant process
computer,
and fuel failures at one foreign
BWR facility was attributed to this
cause.
The impact
on actual
versus
calculated
MCPR values is expected
to be much greater
(about
15%) for reactors
operating
with channels
being
used
in
a
second
bundle lifetime.
The licensee will complete all planned
actions
associated
with this information notice prior to Unit 2 restart.
During this reporting
period
the
inspector
met with members
of
licensee
management
to identify additional
information, if any, that
may
have
been
discovered
by the
licensee
during the fuel
bundle
reconstitution activities that occur red at Browns Ferry during 1988.
The inspector
was informed of the following:
Browns Ferry site
and corporate
standards
do not allow reuse of
fuel
channels
with exposure
greater
than
one
fuel
bundle
lifetime.
This
requirement
has
existed
since initial fuel
loading
and
was also applied to channels
used
on reconstituted
fuel bundles
in that exposure
greater
than
one effective bundle
lifetime would not be exceeded.
No problems with channel
box bowing were identified during the
reconstitution activities,
however
no special effort was
made to
look for evidence of channel
box bow.
The licensee
plans to modify the associated
software to provide
additional
MCPR margin prior to Unit 2 restart.
0
The
NRC inspectors
will follow this
issue
as
part of
a future
inspection associated
with recently issued
NRC Bulletin No. 90-02.
c.
Missed Sample
The inspectors
were informed
on April 5,
1990, that the licensee
had
failed to perform two consecutive
compensatory
samples
on the Unit
1
Raw Cooling Water System.
These
samples
were required
by Technical Specification 3.2.D,
Note
D, at least
every eight hours,
due
the
Radiation Monitor, 1-RM-90-132D,
having
been
declared
on
April 1,
1990.
This failure to sample
was discovered
by licensee
personnel
at
1:50 p.m.
on April 4,
when
upon taking the required
samples it was
determined
that the previous
samples
obtained
at
7:45 a.m. did not identify that the spare
RBCCW Heat Exchanger
sample
was obtained.
After further investigation,
the licensee
determined
that
the
required
sample
at
1:58 a.m.
on April 4
had also
been
missed.
The
licensee
has
determined
that
the failure will be
reportable
to the
NRC in accordance
with
Other,
almost identical,
examples
of failure to perform required
samples
are documented
in LERs 259/88-41,
259/88-51,
and 296-88-06.
Although
this event
was
discovered
by the licensee, it does
not meet
the
criteria for a non-cited violation since the violation is similar to
the violations identified in the
LERs.
Violation'59/90-08-01,
Missed
RCW Samples, will be issued for this violation of Technical
Specification requirements.
One violation
was identified. in the Operational
Safety Verification area.
5.
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 description,
the corrective actions
taken,
the existence
of
potential
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
Momentary
Loss of Secondary
Containment
Caused
by
Failure of Welds
on Door Lock Mechanism.
A breach of secondary
containment
occurred
when personnel
were leaving the
refuel floor to the control building roof and both doors of the airlock
were opened
simultaneously.
This was
caused
by the failure of two welds
which attach
the bracket that holds the lockset in the door on the refuel
floor side of the airlock.
The inspector
reviewed the licensee's
closure
package
for this
LER.
The lockset bracket
was repaired.
Signs
were
placed at the airlocks concerning
proper usage of the doors.
The design
of the
system
interlock was
reviewed
and determined
to
be acceptable.
There
was
no history of similar failures for the doors.
The failure was
caused
by high usage
during the outage..
These actions
were appropriate
to
resolve this item.
This
LER is closed.
6.
Action on Previous
Inspection
Findings
(92701,
92702)
a ~
b.
c ~
(CLOSED) IFI 260/84-41-02,
Stress
Analysis of HPCI Discharge
Pipe.
This
open
item concerned
the failure of Unit 2 HPCI discharge
pipe
supports
R-23
and
R-24 in the
1984 time frame
and the
need for a
stress
analysis
on the associated
piping.
This item was addressed
in
the
Design
Baseline
Verification
Program
and
the
extensive
Bulletin 79-14/02 modifications effort performed
by the
licensee.
The
associated
pipe
stress
problem
N1-273-06R,
Pipe
Support
Calculations
CD-Q2073-883430
(R23)
and
CD-Q2073-891001
(R24),
and
Support
Design
Drawings
2-47B455S0019,(R23)
and
2-478455R0024
(R24)
have
been
issued
to document the system integrity.
While these specific calculations
have-not
been
reviewed
by the
NRC
for closure of this particular item, the
NRC has
conducted
numerous
inspections
of the
DBVP and
IB 79-14/02 engineering calculations
and
has
found the licensee's
programs fully acceptable.
This item is
closed
based
on the calculation
programs
inspection efforts.
(OPEN)
I.FI 259/85-06-02,
IRM Noise
This item was
opened
against Unit 1, but has applicability for all
three units.
It involves erroneous
high reading
on
IRM channels
and
was believed to have
been
caused
by electrical
"cross talk" of IRM,
cables
at containment
For Unit 2 the
SRM and
IRM,
two-shield
coaxial
cables
from the
detector
connector
to
the
preamplifiers
were replaced with a
new improved three-shield
coaxial
cable
as
recommended
by GE-SIL 8192.
The modification was performed
under
5485
and
5534.
The inspectors
have
reviewed the
ECNs,
observed
surveillance,
and
observed
the
SRM operation
during the
recent defueling of Unit 2.
No problems
were identified.
The work
performed will be tracked for Units
1 and
3 modifications,
but this
item is closed for Unit 2.
(CLOSED) IFI 260/85-51-01,
Inspection of Existing Cable Tray Support
Systems.
This item concerns
the fact that in 1985 evidence
could not be found
that
indicated
that
the
cable
tray support
systems
had
been
inspected
to
an
approved
procedure
for verification of as built
condition.'he
seismic qualification of cable tray and cable tray supports
at
Browns Ferry Nuclear Plant Unit 2 was
reviewed
by
NRC as
documented
in the Safety Evaluation
dated
February 5,
1987 (Ref:
NRC Letter,
D.R. Muller
(NRC)
to
S.A.
White
(TVA), "Transmittal
of Safety
Evaluation
Concerning
the Interim Acceptance
Evaluation of Seismic
Qualification of Cable Tray/Supports,"
February
5,
1987).
Issues
related
to cable tray and cable tray supports
are
covered in this
safety evaluation
and are closed
based
on that evaluation for Unit 2
only.
(CLOSED)
IFI 259,
260,
296/86-32-03,
Reactor
Protection
System
Calibration Frequency.
This
item concerned
a discrepancy
between
the safety analysis
which
supported
TS
changes
for the
new
RPS Analog Transmitter
and Trip
Units and actual
plant practice.
The item was reviewed in IR 88-16
and remained
opened
pending resolution of outstanding
discrepancies.
The discrepancies
were that
an
18 month calibration cycle
was not
supportable
for TOBAR transmitters,
and that calculations
for the
calibration
frequency of PT-68-95
and
PT-68-96
were not completed.
The inspector
reviewed
the licensee
closure
package for this item.
TS
amendment
number
167
was
issued
July 7,
1989,
to change
the
calibration
frequency for instrument lines containing transmitters
manufactured
by TOBAR to six month intervals.
The inspector
reviewed
the
TS
change
and
SI,
and
they
had
been
changed
to six month
intervals.
The
inspector
reviewed
the
Setpoint
and
Scaling
Calculation for PT-68-95
and
PT-68-96.
The calculation
compared
the
loop accuracies
to the required accuracies,
setpoints,
safety limits,
and/or operating limits, and concluded
the accuracy of the loops
was
acceptable
for the intended function.
The inspector
concluded that
the
TS changes,
revised SIs,
and calculation resolved
the outstanding
issues
from IR 88-16.
This items is closed.
(Closed)
IFI 259,
260, 296/90-05-01,
ECP Corrective Actions
This IFI identified an example in which
a procedure
change
that was
made
as
a corrective
action
to
a valid
employee
concern
was
subsequently
deleted
from
the
revised
procedure.
The
initiated
CATDs to ensure
that corrective actions
were
implemented
for valid employee
concerns.
In June
1988, licensee
procedures
were
revised
to require
a note with each
procedure
step associated
with
CATO corrective
actions.
The inspector
expressed
concern
that
procedural
corrective
actions
completed prior to June,
1988 could
still
be deleted,
since
they did not include identifying notes.
Licensee
management
stated
that an action plan would be developed to
review this concern.
This item was
opened
to follow the licensee's
actions.
The
inspector
discussed
this
issue
with licensee
personnel
and
reviewed
the licensee's
action
plan
and findings.
The action plan
included
a review of all
CATDs closed
before
October
10,
1988, to
ensure
that procedure
changes
were still in place.
There were
14
restart
CATDs
and
23 non-restart
CATDs identified which included
procedure
changes.
The
licensee
reviewed all of the
procedure
changes
required for the restart
CATDs and
a sample of nine of the
non-restart
CATDs.
In each
instance,
the
changes
were in place in
current
procedures
or
had
been
revised after
subsequent
licensee
reviews.
None of the reviewed corrective actions
implemented
by the
0
CATDs had
been voided.
The licensee
also
added
notes
to several of
the procedure
steps
to indicate that they were associated
with
corrective actions.
The inspector
noted that the action plan was detailed
and the reviews
performed
were indepth.
In addition,
the inspector
noted that the
licensee's
action in response
to this issue
was timely (this issue
was identified during the previous reporting period).
This item is
closed.
(CLOSED) IFI 260/89-20-02 for Unit 2 only,
CRD Seismic Analysis.
The licensee identified an apparent
discrepancy
between
the moment of
inertia (stiffness)
used in
a recent
seismic reanalysis
for the
housings
and
the
moment of inertia
used
in the original stress
evaluation.
This item required
extensive
modifications of the
housing supports
which was followed by the
NRC-Hg staff.
This issue
was identified during the
NRC inspections
performed
from
April 26 to
June
28,
1989
as
reported
in
NRC Inspection
Report
50-260/89-31
dated
July 17,
1989.
The staff
and its consultant
identified three
open issues
in this report.
On
August 14-16,
1989,
the staff
performed
an
inspection
(IR
50-260/89-39
dated
October 13,
1989) to review the resolution of the
open
items
identified
in
IR 89-31.
As
a result of this
NRC
inspection,
two items were still open.
These
two items
were
subsequently
closed
in
NRC
IR 50-260/89-62,
dated
February
16,
1990.
This issue is closed for Unit 2.
(CLOSED)
URI 260/86-06-02, for Unit 2 only.
Reactor
Building Control
Bay
HVAC Inadequate
Design.
This
item
concerned
the
licensee's
identification of inadequate
design
of
HVAC supports.
Interim followup of this
item
was
reported in IR 50-259,
260, 296/89-20,
paragraph
7.C.
The seismic
design of the
HVAC duct and supports
was reviewed
by the
staff in its inspections
of the
BFNP Unit 2 Seismic
Design Program.
As stated
in
NRC Inspection
Report 50-260/88-38,
the staff and its
consultants
identified
several
issues
relating
to this
item.
However,
all
of
these
issues
were
closed
in
subsequent
NRC
inspections.
The following are
the
open
item
numbers
and
the
inspection reports
where these
issues
were closed.
CSG-24
IR 50-260/89-42
dated
February 26,
1990
CSG-29
IR 50-260/89-29
dated
September
20,
1989
CSG-30
IR 50-260/88-38
dated April 19,
1989
0'
The staff has extensively
reviewed this issue
under its inspections
of the Seismic Design Program.
There are
no open items remaining for
this
issue.
This
unresolved
item
is
closed
for Unit
2
based
on
the
above
inspections.
This
licensee-identified
violation
is
not
being
cited
because
criteria
specified
in
V.G.1 were satisfied
.
This item is closed
for Unit 2,
and identified
as
NCV 260/90-08-03,
Reactor
Building
Control
Bay
HVAC Inadequate
Design.
(CLOSED)
URI 260/86-14-03
for Unit
2 only, Overstress
of Drywell
Beams.
This item involves licensee
identified discrepancies
in the drywell
platform design calculations.
These
discrepancies
included:
1)
some
eccentric
loads
were
not included,
2)
some uplift loads
were not
included,
3)
some calculations
were not second
checked,
and 4) the
structural
behavior of the overall platform under combined loads
was
not analyzed.
The structural
evaluation of the drywell steel
platforms were covered
under
the
Browns Ferry Unit 2 Seismic
Design
Program.
During the
inspection of the
TVA calculations for the evaluation of the drywell
steel
platforms,
the staff
and its consultants
identified several
items
as stated
in
NRC Inspection
Report 50-260/88-38
dated April 19,
1989.
These
items
were
numbered
as
CSG-10,
CSG-11,
CSG-12,
and
CSG-14.
All of these
issues
were closed satisfactorily in later
NRC
'inspections.
The following are
the
open
item
number
and
the
inspection report where these
issues
were closed.
CSG-10
CSG-11
CSG-12
CSG-14
IR 50-260/89-42
dated
February
26,
1990
IR 50-260/89-32
dated
November 8,
1989
IR 50-260/89-29
dated
September
20,
1989
IR 50-260/89-21
dated
June
15,
1989
The
staff
has
reviewed
this
issue
under
its
inspections
of the Seismic
Design
Program.
There are
no open
issues
remaining
for this item.
Therefore, this unresolved
item is closed for Unit 2
based
on
the
above
inspections.
This
licensee-identified
violation
is
not
being
cited
becuase
criteria
specified
in
V.G. 1 were satisfied.
This item is closed
for Unit 2,
and
identified
as
NCV 260/90-08-04,
Overstress
of
Drywell Beams.
(CLOSED)
URI 260/87-26-03,
Pump Suction Anchors
and Nozzle
Load
Allowables are Possibly
Exceeded.
This item concerns
RHR load allowables,as
identified by the licensee
in deficiency
number 87-13-6 of Engineering
Assurance
Audit 87-13.
The
licensees
extensive
IB 79-14/02
design
verification
and
e
10
modification program dealt with the specific problem.
The
RHR anchors
and nozzle qualifications
are within the jurisdic-
tional
boundary of the
Long Term Torus Integrity Program
(LTTIP).
These
anchors
serve
as
a boundary
between
the 79-14 stress
problem
Nl-274-9R and the
LTTIP stress
problem Nl-273-5R.
The overlapping
loads
from the 79-14 stress
problem have
been
combined with the LTTIP
pipe stress
problem Nl-273-5R (calculation
CD-f2073-883012).
This
calculation properly documents
the anchor
loads
and the
pump nozzle
qualification.
The pipe support structural
anchors
are within the
LTTIP program.
Because
of the actions
taken
under
these
programs
as part of the
overall
NPP activities, it is not clear that
a violation existed at
the time of IR 87-26.
The efforts of the licensee
and the review
effort by the
NRC staff of the calculation
program
have
addressed
this concern,
and this item is considered
closed for Unit 2.
(CLOSED)
260/85-41-01,
Inadequate
Design
Controls
for
Safety-Related
Cable Tray Supports.
This item concerns:
(1)
Cable tray supports
in the control
bay area
were not seismically
designed.
(2)
Diesel
generator
building cable tray supports
were improperly
designed.
(3)
Cable tray support calculations
in the reactor building showed
lack of thoroughness,
clarity, consistency
and accuracy.
(4)
Design verifications
had not been
implemented
in an acceptable
manner.
The seismic qualification of cable tray and cable tray supports
at
Browns Ferry Nuclear Plant Unit 2 was reviewed
by
NRC as
documented
in the Safety Evaluation
dated
February 5,
1987 (Ref:
NRC Letter,
D.R. Muller
(NRC)
to
S.A.
White
(TVA), "Transmittal of Safety
Evaluation
Concerning
the Interim Acceptance
Evaluation of Seismic
gualification of Cable Tray/Supports,"
February
5,
1987).
Issues
related
to cable tray and cable tray supports
are
covered
in this
safety
evaluation
and
are
closed for Unit
2 only based
on that
evaluation.
7.
Implementation of Nuclear guality Assurance
Plan
(35502)
The inspector
reviewed the status of implemention of the
new
NEAP.
This
plan replaces
the guality Assurance
Program Description
(Topical Report)
TVA-TR75-1A.
Included in this
change
is
a transition
from the current
11
NQAM to
the
Nuclear
Procedures
System.
The
NQAP is to
be fully
implemented
by June 30,
1990.
The licensee
developed
a matrix to show
where
NQAP requirements
are
implemented.
Fourteen
procedures
were
identified that will require
changes.
Each
change
has
been
assigned
a
responsible
site organization for making
the
change.
A schedule
for
completing the
changes
has
been
developed with the last scheduled
change
to
be completed
by June I, 1990.
The
NQAP is described
in TVA document
TVA-NQA-PLN89-A.
High Potential
Testing
of Electrical
Cables
-
Work Observation
and
Procedure
Review (51061,
51063)
The
inspectors
followed ongoing
licensee
activities
associated
with
Special
Test ST-90-01,
Special
Test Procedure
for High Potential
Testing
of
Low Voltage
Cable.
The licensee's
engineering
organization
had
identified the ten conduit that had the greatest possibility for
damage to
cable during "pull bys" These conduit were selected for wet high potential
testing to determine if any cable
damage
could
be detected
that may have
caused
by pull-by cable installation
problems similar to those identified
at
the
Watts
Bar facility.
The inspectors
observed
portions of the
preparations
and setup for the testing,
and actual
high potential testing
for selected
cables
in conduit 3ES-1676-IB.
Most of the cables
included
in this conduit
were multi-conductor cables
routed
from the
3EB,
Shutdown
Board to the Unit
3 Control
Room
Testing
was
directly observed for the following cables:
3ES-2071- IB
The testing
process
consisted
of determinating
both
ends
of the cable
conductors,
injecting tap water into various junction boxes
located in the
Unit 3 Reactor
Building and applying voltages
up to 7200 volts O.C., to
the individual conductors.
The actual lifting and relanding of conductors
was controlled by Work Order 90-02259
and accomplished
in accordance
with
the
requirements
of MAI-3.3, Cable Termination
and Splicing for Cables
Rated
Up to 15000 Volts.
On April 5, during testing
on Cable 3ES-2061-IB,
the
personnel
performing the test
noted
from indications
on the testing
equipment
that the cable
being tested
appeared
to
be shorted.
After
investigating
the problem,
the licensee
determined
that the opposite
end
of the affected
conductor
had not been determinated
at the control
room
panel.
The testing activities were stopped
and
an investigation initiated
to determine
the facts
associated
with the failure to verify that the
conductors
were determinated.
The licensee
determined that two additional
conductors
other
than
the
above
mentioned
conductor
were
also
not
determinated
at the control
room panel.
No evidence of damage
to any
equipment or conductor
has
been attributed to this event.
The inspector
reviewed the work order
and met with licensee
personnel
to
discuss
the event.
The inspector
determined that the work order did not
uniquely identify the specific conductors
to
be lifted, only the
cable'umber.
Cable
had
a total of 12 conductors, all of which were
12
to be tested.
Of the
12 conductors,
seven
were associated
with a single
terminal block and
shown
on
a
common drawing.
Two conductors
were spares,
and
the
remaining
three
conductors
were
associated
with
a
separate
terminal
block.
These
three
conductors
(B11G,
B11R,
B11RG)
were
the
conductors
that
had not been determinated
and are actually identified on
another
drawing,
45N32655-4,
which
had not been
referenced
by the work
order.
The
licensee
personnel
involved
in
the
testing
were
counseled
by
management
and cautioned
on the necessity for attention to detail in the
performance of assigned
duties.
The licensee's
incident critique will be
included in the pre-test briefing for future cables
to
be tested
under
The test director was instructed to personally verify conductor
determinations
prior to performance of future high potential testing.
This failure to maintain
adequate
test control
measures
constitutes
a
violation,
NCV 296/90-08-02,
High Potential
Cable Test Control
Problems,
10 CFR 50 Appendix B, Criterion XI, Test Control.
Due to the fact that
the failure was identified by the licensee
and
prompt corrective action
was
immediately initiated, this failure satisfied
the criteria specified
in Section
V.G. 1 of the
NRC Enforcement Policy for a
NCV.
An
NOV will
not be issued
and
a response will not be necessary.
The inspector
observed
that, for the initial portions of the testing
conducted
until April 5, there
was
no
gA or
gC participation
in the
ongoing activities.
After the testing
resumed
on April 8, the inspector
noted that
a
member of the guality Monitoring Group
was
observing
the
testing activities.
One
NCV was identified concerning
High Potential
Cable
Test
Control
Problems.
Exit Interview (30703)
The inspection
scope
and findings were
summarized
on April 13,
1990 with
those
persons
indicated
in paragraph
1 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
259,260,296/90-,08-01
, 259,260,296/90-08-02
260/90-08-03
260/90-08-04
VIO, Missed
RCW Samples,
paragraph
4
NCV, High Potential
Cable Test Control
Problems,
paragraph
8
Reactor
Building Control
Bay
Inadequate
Design,
paragraph
6.g.
Overstress
of
Drywell
Beams,
paragraph
6.h.
13
BFNP
CAQR
CATD
CFR
DBVP
HQ
IB
IFI
IR
KV
LCO
LER
LTTIP
MAI
NQAM
NQAP
NRC
RCW
SDSP
SFRC
TS
Browns Ferry Nuclear Plant
Boiler Water Reactor
Corrective Action Report
Corrective Action Tracking Document
Code of Federal
Regulations
Control
Rod Drive System
Design Baseline
and Verification Program
Engineering
Change Notice
Employee Concerns
Program
High Pressure
Coolant Inspection
Headquarters
Heat, Ventilation,
8 Air Conditioning
Enforcement Bulleting
Inspection
Followup Item
Inspection
Report
Intermediate
Range Monitor
Kilovolt
Limiting Condition of Operation
licensee
Event Report
Long Term Torus Integrity Program
Modification Alteration Instruction
Minimum Critical Power Ratio
Maintenance
Request
Non Cited Violation
Nuclear Performance
Plan
Nuclear Quality Assurance
Manual
Nuclear Quality Assurance
Plan
Nuclear Regulatory Commission
Pressure
Transmitter
Quality Assurance
Quality Control
Reactor Building Closed Cooling Water
Raw Cooling Water
Residual
Heat Removal
Reactor Protection
System
Site Director Standard
Practice
Frequency
Reduction Coordinator
Frequency
Reduction
Program
Surveillance Instruction
Service Information Letter
Source
Range Monitor
Special
Test
Technical Specification
Valley Authority
Unresolved
Item
Violation
Work Order
Work Request