ML18029A670
| ML18029A670 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 06/21/1985 |
| From: | Brocks C, Cantrell F, Patterson C, Paulk G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18029A667 | List: |
| References | |
| 50-259-85-28, 50-260-85-28, 50-296-85-28, NUDOCS 8507110348 | |
| Download: ML18029A670 (14) | |
See also: IR 05000259/1985028
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323
Report Nos.:
50-259/85-28,
50-260/85-28,
and 50-296/85-28
Licensee:
Valley Authority
500A Chestnut Street
Chattanooga,
TN
37401
Docket Nos.:
50-259,
50-260
and 50-296
License Nos.:
and
Facility Name:
Browns Ferry 1, 2,
and
3
Inspection
Conducted:
April 26 - Nay 25,
1985
Inspectors:
. L.
au
,
endor
R
s ent
C.
. Patt
son,
Reside t
D
e
igne
Da
e
igned
0
. Bro
s,
Resi
en
Approved by:
F.
.
antre
,
ecti
,
e
Division of Reactor
Pr
ects
a
e
igne
g zrsg
ate
sgne
SUNNARY
Scope:
This routine,
unannounced
inspection entailed
195 inspector-hours
in the
areas
of operational
safety,
maintenance
observation,
reportable
occurrences,
onsite review committee,
and surveillance
observation.
Results:
VIOLATIONS-
(1)
two examples
of failure to submit licensee
ever/ reports.
(2)
10 CFR 50 Appendix B, Criterion XVI:
failure to take corrective
action to determine root cause
of repeated
SLC system
transformer failures.
(3)
Technical Specification 6.3.A.:
two
examples
of inadequate
procedures:
(a) fire protection
surveillance
inadequate
in
checking
smoke
detectors
in
accordance
with manufacturer's
instructions; fire protection surveillance
inadequate
in verifying
principal
valves
open
quarterly,
and
(b)
Operating
8507ii0348 850624
ADOCK 05000259
Q
Instruction
71 for
RCIC system
inadequate
in verifying proper
valve lineup for all system valves.
(4)
10 CFR 50 Appendix B, Criterion V:
one example of failure to meet
design
specification
with respect
to
emergency
battery
rack
installations
and
one
example of failure to meet workplan review
requirements.
(5)
Technical Specification 6.2.B.4.e.:
failure of PORC to review two
significant events
affecting the
abnormal
performance
of plant
equipment.
(6)
10 CFR 50, Appendix B, Criterion II: failure to conduct
a
10 CFR 21 review in accordance
with plant requirements.
(7)
Technical Specification 3.11.A.1:
failure to
have
two high
pressure fire pumps aligned to the high pressure fire header.
REPORT
DETAILS
Persons
Contacted
Licensee
Employees
J.
A. Coffey, Site Director
G. T. Jones,
Plant Manager
J.
E. Swindell, Superintendent
- Operations/Engineering
J.
R. Pittman, superintendent
- Maintenance
J.
H. Rinne, Modifications Manager
J.
D. Carlson, guality Engineering Supervisor
D. C. Nims, Engineering
Group Supervisor
R. Hunkapillar, Operations
Group Supervisor
C.
G. Wages,
Mechanical
Maintenance
Supervisor
T.
D. Cosby, Electrical Maintenance
Supervisor
R.
E. Burns, Instrument Maintenance
Supervisor
A. W. Sorrell, Health Physics
Supervisor
R.
E. Jackson,
Chief Public Safety
T. L. Chinn, Senior Shift Manager
T. F. Ziegler, Site Services
Manager
J.
R. Clark, Chemical Unit Supervisor
B. C. Morris, Plant Compliance Supervisor
A. L. Burnette, Assistant Operations
Group Supervisor
R. R. Smallwood, Assistant Operations
Group Supervisor
T.
W. Jordan, Assistant Operations
Group Supervisor
S.
R. Maehr, Planning/Scheduling
Supervisor
G.
R. Hall, Design Services
Manager
W. C. Thomison, Engineering Section Supervisor
A. L. Clement,
Radwaste
Group Controller
R. L. Lewis, Senior Shift Manager
Other
licensee
employees
contacted
included
licensed
reactor
operators,
auxiliary operators,
craftsmen,
technicians,
public safety officers, guality
Assurance,
design
and engineering
personnel.
Exit Interview (30703)
The inspection
scope
and findings were summarized
on Nay 23, 1985, with the
Plant
Manager =and/or Assistant
Plant
Managers
and other
members
of his
staff.
The licensee
acknowledged
the findings and took no exceptions.
3.
Licensee Action on Previous
Enforcement Matters
(92702)
a
~
b.
c ~
(Closed)
Open
Item (259/82-34-04)
The inspector
reviewed the as-built
drawing for ultrasonic examination calibration block BF-16.
This item
is closed.
(Closed)
Violation (259,
260,
296/82-34-06)
The inspector
reviewed
Standard
Practice
BF
17.19,
program
to
Establish
and
Maintain
Certifiably Accurate
Thermometers,
and
a similar violation 259, 260,
296/85-06-03.
Violation 82-34-065 is closed
and further tracking in
this area will be under 85-06-03.
(Closed)
Open
Item (259, 260, 296/82-37-02)
The licensee
has
changed
the relief valves
on the portabTe accumulator nitrogen charging
system.
A charging
system
was
examined
and
a relief valve with a setpoint of
1150 psig
was installed.
The licensee
is administratively controlling
the installation of the
new relief valves
under
a temporary alteration
control
form (0-84-040-85)
pending
update of the drawings.
This item
is closed.
d.
e.
(Closed)
Open
Item (259,
260, 296/82-13-02)
The inspector
reviewed
control
diagram
and
the
same
drawing in
FSAR figure
10.11.la
and in both cases
Note
6 was deleted.
This item is closed.
(Closed)
Open Item (259/82-46-02),
Control
Room Emergency Ventilation
System
HEPA Filter Test,
Surveillance
Instruction S.I. 4.7.E.3,
was
reviewed for updating
and clarification.
The S. I. has
been revised at
least
11
times
since
the
82-46
inspection
and
a
complete
update
performed
on
2/12/85.
A cross-check
between
the
S. I.
and
the
implementing
Technical
Instruction
17B,
Freon II Test for Charcoal
Absorber Banks, revealed
no discrepancies.
This item is closed.
(Closed)
Open
Item (259,
260, 296/82-34-07)
Surveillance Instruction
S.I. 4.8.A. 1, Release
Procedure
- Liquid Effluents,
has
been revised to
require addition of nitric acid to prevent absorption.
This item is
closed.
0
g
(Closed)
Open
Item (259/85-06-10)
This item was left open to follow
the licensee's
evaluation of false indication of Main Steam Relief
Valve
(MSRV) position
by the acoustic
valve flow monitor during
a
reactor
on Unit I, January
16,
1985.
The licensee's
failure
investigation report
number
FI 85-10
Rev.
1 with an attached
letter
from the
manufacturer,
dated
February
12,
1985,
indicates
that
a
combination of the signal strength
from the sensor
plus the module gain
can result in significant overdriving of a bar graph driver chip when
an
MSRV lifts such that the bar graph light emitting diodes
(LEDs) will
indicate
flow when the
MSRV subsequently
closes.
The manufacturer
stated that
he changed
to
a different driver chip about
18 months
ago
that
does
not exhibit the "latch-up" characteristic.
The licensee
had
failed to evaluate this concern for reportability under
10 CFR 21 as of
May 16, 1985.
This is
a violation for failure to carry out the quality
h.
assurance
program
in accordance
with the written
procedures
and
instructions
governing
10 CFR 21 reportabi lity evaluation.
(259, 260,
296/85-28-01)
and
was
discussed
with the licensee
during
an exit on
May 23,
1985.
During the course of this follow-up, it was discovered
that the licensee
had failed to submit an
LER within 30 days describing
the
event
during which the
acoustic
monitor
problem
became
evident.
The
scram occurred
on January
16,
1985 and contained
several
automatic
ESF actuations "including the Reactor Protective
System
(RPS),
High Pressure
Coolant
Injection
System
and
Reactor
Core Isolation
Cooling System
(RCIC).
This violation 259/85-28-02)
was also discussed
during
the exit
as
the
second
example
of the
violation.
(Closed)
Open
Item (259/85-06-04)
This item was left open to review
the licensee's
evaluation of recurring failures of the Unit
1 Standby
Liquid Control
(SLC)
pump
suction
trace
heater
transformer.
The
licensee's
fai lure investigation
report
No.
FI 85-18 attributed
the
fai lure to an overload of the 0.5
KVA transformer
by about
39K.
A
discrepancy
in
the
vendor
(Nelson
Electric)
supplied
drawings
apparently
resulted
in the overload condition which was reported
to
have
been
in existence
since original construction
in 1972.
The
inspectors
obtained
a
power
stores
transaction
history for the
transformer
and found evidence of an excessive failure history for the
Unit 1
SLC transformers
as
summarized
below:
Date
1/26/79
3/01/80
2/15/81
9/22/83
1/13/84
1/11/85
Usage
18
SLC pump trace heater
lA SLC pump trace heater
SLC trace heating
1A SLC pump trace heater
1A SLC pump trace heater
1A SLC pump trace heater
This item was identified as
a violation (259, 260, 296/85-28-03)
during
an exit meeting
on Nay 23,
1985, for fai lure to promptly identify and
correct
the
cause
of repeated
failures.
An additional
violation
(259/85-28-02)
was
discussed
relating
to the failure to
submit
a
Licensee
Event Report
(LER) within the required 30-day time period once
the failure investigation discovered
that overload circuits existed in
both
the =normal
and
redundant
SLC heat
traces
for Unit 1.
This
violation is the first example of the violation of 10 CFR 50.73(a)(2).
4.
Unresolved
Items* (92701)
There
was
one
unresolved
item
(259,
260,
296/85-28-09)
during this
inspection.
On Nay 20,
1985,
the licensee
discovered that
some
Secondary
Containment
blowout panels
which protect the reactor building against
n
nreso
ve
tern
ss
a matter
a out which more information is required
to
determine whether it is acceptable
or may involve a violation or deviation.
excessive differential pressures
during tornadoes
and
steam line breaks
had
not
been maintained
in accordance
with design
drawings.
The licensee
is
evaluating this event to determine
when apparent modifications to the panels
were
made which rendered
them incapable of blowing out at their required
36
pounds
per
square
foot differential pressure.
Also being tracked in this
unresolved
item is the ineffectiveness
of Mechanical
Maintenance
Instruction
14,
Inspection
of Secondary
Containment Relief Panels,
in detecting this
problem earlier.
Operational
Safety (71707,
71710)
The inspectors
were kept informed
on
a daily basis
of the overall plant
status
and
any significant safety matters
related
to plant operations.
Daily discussions
were held
each
mor'ning with plant management
and various
members of the plant operating staff.
The inspectors
made frequent visits to the control
rooms
such that each
was
visited at least daily when an inspector
was
on site.
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
tempor'ary
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 their supervisors.
General
plant tours
were conducted
on at least
a weekly basis.
Portions of
the turbine building, each reactor building and outside
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;
radiation
protection
controls
adequate;
vital area
controls;
personnel
search
and
escort;
and vehicle search
and escort.
Informal discussions
were held with
selected
plant
personnel
in their functional
areas
during
these
tours.
Weekly verifications of system status
which included major flow path valve
alignment,
instrument
alignment,
and
switch
position
alignments
were
performed
on the control rod drive hydraulic systems.
A complete
walkdown of the accessible
portions of the Unit 3 Reactor
Core
Isolation
Cooling
System
(RCIC)
was
performed
to confirm system
lineup
procedures
match
plant
drawings
and
the as-built configuration
and to
identify
equipment
conditions
that
might
degrade
system
performance
(hangers,
supports,
housekeeping,
etc.).
The following discrepancies
were
noted
and
discussed
with licensee
representatives
during
a routine daily
meeting
on May 10,
1985:
a.
Nuts were missing
on
a pipe clamp near valve 3FCV71-37.
Valve 3-71-524
was missing
an identification tag.
The
RICI
pump discharge
flow element
(PE-71-36)
was leaking drops of
water
form the orifice flange.
A Structural
"I" beam
above
the
RCIC condensate
drain valves
was found
to have nuts about
one inch away from being snug.
Electrical conduit to two
RCIC components
was being supported
by rope
and baling wire.
The valve checklist contained in Operating Instruction 71, Reactor
Core
Isolation Cooling System,
erroneously lists valves
71-216A and 71-217A
as
the root isolation valve fof'team flow instrument
PDIS-71-lA and
PDIS-71-1B.
The drawings
and walkdown confirmed that valves71-221
and
71-222 are the root isolation valves for PDIS-71-1B.
This instrument
initiates
RCIC isolation in the event of a
RCIC steam line rupture.
This item was identified
as
the
second
example of a violation (259,
260,
296/85-28-04) for inadequate
procedures
required
by T.S.
6..3.A.
during an exit meeting
on May 23, 1985.
During
a routine inspection
on April 5,
1985,
the resident
inspector
noted that the Unit 1 and
2 diesel
generator
(D/G) battery A, B, C, and
D racks,
and the unit 3 D/G battery A, B, C, and
D racks were installed
out of drawing specification
(Reference
I.E.
Report
85-25).
On
April 22,
1985,
on
a followup inspection,
the inspector
noted that the
3EB
shutdown
board battery
rack
had
a 3-inch
gap
between
the rack
supports
and the
end battery cell.
Licensee evaluation indicated that
the
D/G battery configuration
and
the. 3EB
shutdown
board
battery
configuration
did not meet
seismic qualification requirements.
The
safety
evaluation
showed
there
was
the potential
to
damage
the
3EB
shutdown
board battery
and
the
DG batteries
during the design
basis
The
3EB shutdown
board battery provides control
power to
the
3EB 4160-V shutdown
board.
The loss of one
4160-V shutdown board
is
acceptable
per
the
Final
Safety Analysis
Report
and
would not
jeopardize
plant safety.
Damage
to the
DG batteries
would prevent
startup of the corresponding
DG.
The loss of the
DG would accordingly
jeopardize
the ability to maintain
the plant in
a
safe
shutdown
condition in the event of concurrent
loss of offsite power (Reference
The diesel
generator
racks
and the
3EB shutdown
board battery rack have
been modified by the licensee to meet seismic requirements.
This item was identified to the Plant Nanager at the exit on Nay 23,
1985,
as
the first example of a violation of
Criterion V, for failure to have battery
racks installed
per design
drawings.
(259, 260, 296/85-28-05).
On
May 23,
1985,
the
High
Pressure
Fire Protection
System
was
discovered
to
be
due
to
a
valve misalignment
problem
following maintenance.
Plant personnel
made
a one-hour report to the
NRC Operations
Center at
2229
on
May 23,
1985,
which 'described
the
violation of Technical
Specification limiting condition 3.11.A.
Fire
pump isolation valves 0-26-527
and
529 were closed
on May 20, 1985, for
maintenance
on a leaking valve.
Since this condition removes all three
fire
pumps
from service,
a
special
operating
instruction
was
implemented
which
placed
the
emergency
diesel
driven fire
pump in
service
and
stationed
the
necessary
roving fire watches.
Upon
completion of the maintenance,
the hold order tag out (H.O,85-688)
was
released,
valves realigned
and the
A fire pump was declared
operable at
1530 on May 21,
1985.
The fire watches
were also secured at this time.
Due to
a training
and procedural
deficiency
on valve operation,
valves
0-26-527
and 0-26-529
remained
shut until 2130 on May 23, 1985.
During
this time, the plant violated Technical Specification
LCO 3.11.A (259,
260, 296/85-28-10).
A similar event involving improper valve operation
of the High Pressure
Fire Protection
system is discussed
in violation
296/81-18-07.
6.
Maintenance
Observation
(62703)
Plant
maintenance
activities
of 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
accom-
plished
by qualified personnel;
parts
and materials
used
were properly
certified;
proper
tagout
clearance
procedures
were
adhered
to; Technical
Specification
adherence;
and radiological
controls
were
implemented
as
required.
Maintenance
requests
were reviewed to determine
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
below listed maintenance activities during this report period:
a.
NNI 22, Reactor
Core Isolation Cooling System,
Section
2B, Lubrication
System Maintenance.
b.
NMI 141, Lubrication of Equipment.
c.
NMI 14, Inspection of Secondary
Containment Relief Panels.
d.
Diesel generator fuel line filter support repairs - Unit 3.
Surveillance Testing Observation
(61726)
The
inspectors
observed
and/or
reviewed
the
below listed surveillance
procedures.
The inspection
consisted
of a review of the
procedures
for
technical
adequacy,
conformance
to technical specifications, verification of
test instrument calibration, observation
on the conduct of the test,
removal
from service
and return to service of the system,
a review of test data,
limiting condition of operation
met,
testing
accomplished
by qualified
personnel,
and
that
the
surveillance
was
completed
at
the
required
frequency.
A-S.I. 4.11.A.5
B-S.I. 4.11.C.1
and C.5
High
Pressure
Fire
Protection
System
Valve
Alignment
Testing of Smoke
and Heat Detectors
C-S.I. 4.7.E
D. 0.1.
85
Control
Room Emergency Ventilation System Testing
Control
Rod Drive Hydraulic System
S.!. 4.11.A.5 is intended
to satisfy Technical
Specification surveillance
requirement 4.11.A.5 that principal header
and component isolation valves of
the high pressure fire protection
system
be checked
open quarterly.
The SI
was
found to be deficient in that several
major valves were not included
on
the SI checklist.
These
include
0-26-1400,
0-26-1401
isolations
south
side
from
emergency
diesel-driven
fire
pump),
1-26-2397,
1-26-1363,
1-26-1364,
3-26-1398,
3-26-1367,
3-26-1368
(Units 1, 2,
and
3 cable
spreading
room
preaction
sprinkler
system isolation valves).
This is
an
example of a
violation against
Technical Specification 6.3.A.
and
was
discussed
with
licensee
representatives
during the inspection.
It was
learned that the
licensee
was
aware of drawing
and
procedure
problems with the
H.P. Fire
Protection
System
and that
a complete
system
walkdown has
been
performed
by
plant personnel.
Work plan No. 0049-84
had been prepared
to correct drawing
and
procedural
deficiencies
and
was
nearly
completed.
The
inspector
reviewed this workplan
and
noted that plant personnel
had failed to list
S. I. 4.11.A.5
as
one of the Instructions requiring revision
as
a result of
the
walkdown.
Standard
Practice
8.3,
Plant Modifications,
governs
the
workplan prep@ation
and
requires
in section
8.3. 1 that affected
work-
sections list any plant instructions
requiring review and updating in the
workplan.
Due to this oversight,
the workplan would not have instituted the
required
changes
to S. I. 4.11.A.5 to correct the deficiencies.
This is the
example of a failure to adhere to procedures
required
by 10 CFR 50 Appendix
B, Criterion V. (259, 260, 296/85-28-06).
S. I.
4. 11.C.1
and
C.5 is
intended
to satisfy
Technical
Specification
surveillance
requirement 4.11.C.5 that
smoke detector sensitivity be checked
in accordance
with manufacturer's
instructions.
The inspector obtained the
manufacturer's
instructions for the Fire Alert Model
CPD-1212
smoke detector
s.
and the Fire Alert Model
FT-200 smoke detector
and found that S.I. 4.11.C.l
and
C.5
was deficient.
Walter Kidde 5
Company Bulletin 841 "Fire Alert
CPD-1212 Installation
and
Technical
Data"
requires
an
adjustment
of the
sensitivity control to maximum and then minimum with an acceptance criteria
of not less
than 0.6 volts (on maximum sensitivity)
and not less
than 1.1
volts
(on minimum sensitivity).
Section 4.2.1.2 of S. I. 4. 11.C.1
and
C.5
does
not specify
any setting of the sensitivity control
and contains
an
acceptance
criteria of not less
than 0.6 volts or greater
than 2.0 volts.
Should sensitivity control
be set
a minimum (as sent from the factory), the
S. I. would accept
a sensitivity
as
low as
0.6 volts contrary to the
manufacturer's
criteria of not less
than
1. 1. volts.
This
item
was
identified
as
an
example of an inadequate
procedures
violation during the
exit meeting
on May 23,
1985.
The Surveillance Instruction was additionally
inadequate
in its method of sensitivity testing
on the Fire Alert Model
FT-200
smoke detectors.
The instruction in S. I. 4. 11.C. 1 and C.5 relies
on
a determination of the flash rate from a light emittimg diode
(LED) on the
detector.
No manufacturer's
instructions
could
be located
by the licensee
to support this method.
Walter
Kidde Bulletin 911 "Fire Alert FT-200
Ionization
Smoke Detectors" describes
the use of an
FTM-200 Field Test Meter
which plugs
in series
with the detector
to provide
a
means for testing
detector sensitivity (259, 260, 296/85-28-04).
Reportable
Occurrences
(90712,
92700)
The below listed licensee
events
reports
(LERs) were reviewed to determine
if the
information
provided
met
NRC requirements.
The determination
included:
adequacy
of event description, verification of compliance with
technical
specifications
and
regulatory
requirements,
corrective
action
taken,
existence
of potential
generic
proNems,
reporting
requirements
satisfied,
and the relative safety significance of each event.
Additional
in-plant reviews
and discussion
with plant personnel,
as appropriate,
were
conducted
for those
reports
indicated
by
an asterisk.
The following
licensee
event reports
are closed:
LER No.
- 296/85-02
- 296/85-05
- 296/85-06
- 296/85-08
- 296/85-10
Date
2-05-85
2-10-85
2-13-85
3-09-85
4-21-85
Event
Manual Scram Unit 3
in Unit 3 drywell
Reactor water level instrument mismatch
due to rod worth minimizer
system inoperability
Inadvertent
initiation of
containment
isolation system
No violations or deviations
were found in the above.
9.
Regulatory Performance
Improvement
Program
(RPIP)
The resident
inspector
attended
an employee
involvement session
led by the
site director
on
Nay 8,
1985.
The responsible
Region II section chief
reviewed the status
of RPIP
and actions
taken
by TVA to implement specific
items
as required
by
NRC Confirmatory Order
EA 84-34 dated July 13,
1984.
TVA has
assigned
a senior
manager
as
RPIP Coordinator
at the site.
His
responsibilities
include verifying that
each
task
has
been
implemented
as
described,
has
met objectives,
and that the necessary
programs
are in place
to insure that objectives will continue to be met.
The status of RPIP and
action taken
by TVA to implement specific items
as required
by
NRC Confirma-
tory Order
EA 84-34
was reviewed during
a site visit Nay 28-30,
1985.
All
of the short term items
have
been accepted
as complete
by TVA.
This review
included discussions
with responsible
personnel
and records of action taken
to close the specific items.
Based
on the above review the following items
are closed:
Short Term
Item Number
2.1
(84-SC-07)
2.6
(84-SC-12)
2.7
(84-SC-13)
2.9
(84-SC-15)
3.1
(84-SC-18)
3.11
4.3
(84-SC-30)
Reduce level of activity at plant site.
Establish
and train planning group.
Revise modification procedures
to
eliminate unnecessary
reviews
and
better define responsibilities.
Redefine priorities for U-3 modification
work.
Provide statement
of reviewers
responsibility.
Ensure all employees
understand
RPIP.
Provide facilities for new organization.
Additional permanent facilities being
provided and tracked under long term item
9.13.
4.6
(84-SC-32)
4.7
(84-SC-33)
Implement Site Director's Office.
Implement Technical
Services.
10
4.8
(84-SC-34)
4,9
(84-SC-35)
4.10
(84-SC-36)
4.11
(84-SC-37)
4. 12
(84-SC-38)
4.13
(84-SC-39)
4.14
(84-SC-40)
4.15
(84-SC-41)
4.16
(84-SC-42)
4.20
(84-SC-44)
4.21
(84-SC-45)
4.22
(84-SC-46)
4.23
(84-SC-47)
4.24
(84-SC-48)
4.25
(84-SC-49)
4.26
(84-SC-50)
5.0
(84-SC-52)
Implement Administration.
Implement Modification Group.
Implement Design Services.
Implement Independent
Safety Engineering.
Implement Personnel
Office.
Implement Plant Manager's Office.
Implement Engineering Section.
Implement Other Functions.
Implement Operations
Group.
Implement Public Safety Section.
Implement Planning
5 Scheduling.
Implement Compliance Staff.
Implement Mechanical
Maintenance
(Moved to Long Term 9.16).
Implement Electrical Maintenance
(Moved to Long Term 9.17).
Implement Instrument Maintenance.
Implement Training Organization.
Reassign/retrain
based
on performance.
11
6.0
(84-SC-53)
7.1
(84-SC-54)
Assign division program manager
on site.
Place responsibility for licensed
operator training under Nuclear Training
Bra'nch.
7.2
(84-SC-55)
7.3
(84-SC-56)
The following short term items
are
action planned
by TVA:
Short Term
Item Number
Institute accelerated
retraining.
Provide "live time" training on shift.
open
pending verification of additional
Oescri tion
3.2
(84-SC-19)
Assign modification on project basis,
(RPIP) manager
needs
to review consul-
tants'eport,
make recommendations
and
develop
program to implement
recommen-
dation.
3.4
(84-SC-21)
3.5
(84-SC-22)
3.7
(84-SC-24)
3.8
(84-SC-25)
4.5
(84-SC-31)
Continue with Procedure Training.
Records
reviewed
showed the number of
craft personnel
that had received
training, but not total
number requiring
training.
Assign system responsibility to section
engineers
and provide training (Basis
for closing not clear - additional
action needed
to establish
continuing
program).
Provide training in preparation of
safety evaluation
technique
(Additional
action required to establish
continuing
program).
Ensure
each
employee
understands
his
responsibility for quality and
compliance
(Additional action required
to establish
continuing program).
Revise appropriate
documents
to define
new organization
(Additional action
require with respect to technical
specification
and the security plan).
12
4. 19
(84-SC-43)
4.27
(84-SC-51)
Onsite
Review Committee
(40700)
Implement Health Physics Organization
(Review required
by Region II Radiation
Protection Group).
Implement
gA Organization
(Action
required to fill key positions, i.e.,
need operations qualified personnel).
The Plant Operations
Review Committee
(PORC)
was inspected
to verify that
was
being satisfied
regarding
membership,
meeting
frequency,
duties,
responsibilities,
and
records.
The resident
inspector
attended
a routinely scheduled
PORC meeting
on May 14, 1985,
as
a
nonparticipating
observer.
meeting
minutes for the
period
from
December
1984 through March 1985 were reviewed.
Although T.S.
6.2.B. 1 specifies
the plant superintendent
will serve
as
Chairman of the
PORC,
PORC meeting minutes for the period of December
1984
through March 1985 indicate that he routinely delegates
this responsibility.
In fact
he did not Chair
any
PORC meetings
during this period.
The
minutes for at least five meetings
(12/16/84, 1/17/85, 1/18/85, 3/ll/85 and
3/12/85) listed the plant superintendent
as
being in attendance,
however,
his representative
Chaired
the meeting.
This does
not appear
to meet the
intent of T.S.
6.2.B.1
which only allows
a representative
to serve
as
Chairman in the absence
of the plant superintendent.
This was identified as
an open item (259/85-85-28-07)
during an exit meeting
on May 23,
1985.
The duties
and responsibilities
of the
PORC as listed in T.S. 6.2.8.4.e
requires
the
PORC to review reportable
events,
unusual
events,
operating
anomalies
and abnormal
performance of plant equipment.
No records
could be
located to confirm that the
PORC reviewed
two events
which resulted
from
abnormal
performance
of plant
equipment
that
were classified
as
a
Notification of Unusual
Event
(NOUE) by plant personnel
on January 9,
1985
and
on
January
11,
1985.
The
on
January
9,
1985
involved the
inoperability of the containment
cooling
mode of the Residual
Heat
Removal
(RHR) System
on Unit 1
(
a one-hour report per
10 CFR 50.72(b)( 1) was
made
by the licensee
during the event).
The
NOUE on January
11,
1985 involved
the inoperability of the Standby Liquid Control
System
on Unit 1.
This item
was identified
as
a violation (259,
260,
296/85-28-08)
during
an exit
meeting
on May 23, 1985.