ML18030B171
| ML18030B171 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/11/1986 |
| From: | Brooks C, Cantrell F, Patterson C, Paulk G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18030B170 | List: |
| References | |
| 50-259-86-05, 50-259-86-5, 50-260-86-05, 50-260-86-5, 50-296-86-05, 50-296-86-5, NUDOCS 8603110551 | |
| Download: ML18030B171 (28) | |
See also: IR 05000259/1986005
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323
Report Nos.:
50-2S9/86-05,
50-260/86-05,
and SO-296/86-05
Licensee:
Valley Authority
500A Chestnut Street
Tower II
Chattanooga,
37401
Docket Nos.:
50-259,
50-260,
and 50-296
License Nos.:
and
Facility Name:
Browns Ferry Nuclear Plant
Inspection
Conducted:
January
1-31,
1986
Inspectors
G.
L.
Pa
k, Senior
Resi
nt
C. A. Pat
rson,
Resid nt
4io
Date Signed
]o 3'(
Dat
Signed
C.
R,
Br
ks, Resident
Approved by:
.
S. Cantrell,
S
on Chief
Division of React r Projects
Dat
Signed
Date Signed
SUMMARY
Scope:
This routine
inspection
involved
280 resident
inspector-hours
in
the
areas
of operational
safety,
maintenance
observation,
reportable
occurrences,
surveillance,
and calibration.
Results:
One Violation -
10 CFR 50, Appendix B, Criterion III for failure to
reference
the correct design specification
on design drawings.
8603110551
8603
>~59
pDR
ADOCK 0500p~~R59
9
REPORT
DETAILS
Licensee
Employees
Contacted:
W.
C. Bibb, Site Director
T.
F. Ziegler, Assistant to the Site Director
R.
L. Lewi s,
Pl ant Manager
E. A. Grimm, Assistant to the Plant Manager
J.
E. Swindell, Superintendent
Operations/Engineering
T.
D. Cosby,
Superintendent
Maintenance
J.
HE Rinne, Modifications Manager
D.
C. Nims, Engineering
Group Supervisor
R.
M. McKeon, Operations
Group Supervisor
C.
G.
Wages,
Mechanical
Maintenance
Supervisor
J.
C. Crowell, Electrical Maintenance
Supervisor
R.
E. Burns,
Instrument Maintenance
Supervisor
A. W. Sorrell; Health Physics
Supervisor
R.
E. Jackson,
Chief Public Safety
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
S.
R. Maehr, Planning/Scheduling
Supervisor
W.
C. Thomison,
Engineering
Section Supervisor
C.
E. Burke,
Radwaste
Group Controller
Other
licensee
employees
contacted
included
licensed
reactor
operators,
auxiliary operators,
craftsmen,
technicians
public safety officers, (}uality
Assurance,
Design
and engineering
personnel.
Retired Admiral Steven
A. White,
who has
served
as Chief of Naval Material
and
as
Commander,
Submarine
Force,
U.S. Atlantic Fleet,
assumed
the position
of Manager of the
TVA Office of Nuclear
Power effective January
13,
1986.
The
TVA Board of Directors
announced
that it contracted with Stone
5 Webster
Engineering
Corporation
of Boston,
for
the
assignment
of
White to provide direct management
of TVA's nuclear
power program.
White's
services
are
being
provided to Stone
8 Webster
by Stemar
Corporation,
of
which
White is
the
principal
officer.
This
management
arrangement
is
intended
to bring additional
top nuclear
experience
on board quickly and
expedite
the filling of key positions in the nuclear
program with permanent
TVA employees.
As Manager
of Nuclear
Power,
White will report to the
General
Manager
and Board of Directors.
Exit Interview
(30703)
The inspection
scope
and findings were
summarized
on February 3,
1986, with
the Plant Manager and/or Assistant Plant Managers
and other
members
of his
staff.
The licensee
acknowledged
the findings and took no exceptions.
The licensee
did not identify as proprietary
any of the materials
provided to or reviewed
by the inspectors
during this inspection.
Licensee Action on Previous
Enforcement Matters (92702)
(Closed) Violation (259/260/296/85-15-04)
Standard
Practice
12.7 for shift
turnover and 7.6 for maintenance
requests
were reviewed for revisions which
should provide better tracking of inoperable
equipment.
Monitoring of the
standby
gas
treatment
humidity heater
breaker
currents
over
several
months
has revealed
no cause for the breaker tripping'.
This item is closed.
(Closed)
Open
Item (259/260/296/85-15-05)
The licensee
received
from the
vendor time curves for the
standby
gas
treatment
system relative humidity
breakers
and
has
evaluated
the breakers
in question.
Testing
revealed
no
.
noted deficiencies with the breakers.
This item is
closed'Closed)
Violation
(259/260/296/85-09-01)
Operating
Instruction
OI-30,
Ventilation
System,
was revised
March 8,
1985, to address
concerns
about the
ventilation
system
lineup.
Surveillance
Instruction,
SI 4.2. F. 18,
was
revised
March 5,
1985 to provide
acceptance
criteria for the
main
steam
relief valve tailpipe thermocouples.
This item is closed.
(Closed)
Open
Item (259/85-32-03)
Thi's i'tern concerning
seismic qualification
of the reactor
building overhead
crane
wi 11
be tracked under'he
Licensee
Event Report 259/85-27 which has not been closed.
(Closed)
Open Item (259/85-32-02)
This item concerning
seismic qualification
of the fuel
pool cooling
pump flanges
due <o raised
faces will be
tracked
under the Licensee
Event Report 259/85-27 which has not been closed.
(Closed)
Open
Item (259/85-32-01)
This item concerning deficiencies with the
250
VDC system will be tracked
under
the
Licensee
Event
Report
259/85-32
which has not been closed.
(Closed)
Violation
(259/85-28-03)
The
inspector
reviewed
the
failure
trending
program,
applicable
section
Instruction
Letter
EMSIL 35,
and
discussed
the
program with the responsibl~
engineer.
This program
should
provide timely detection
and correction of repetitive equipment failures in
the future.
This item is closed.
(Closed)
Violation (259/85-28-02)
The
inspector
reviewed
the
response
to
this violation for fai lure to submit Licensee
Event Reports
(LERs).
Reports
259/85-47
and 259/85-16
have
been
submitted.
This item is closed.
(Closed)
Violation
(259/260/296/85-28-01)
Both
the
LER
revision
for
259/85-16
related
to the
main
steam relief valve acoustic
monitor
and
Standard
Practice
6. 18 concerning failure investigation
were
reviewed for
completion of the licensee's
corrective action.
These
items
were
complete
and the inspector
has
no further questions.
S
(Closed)
Unresolved
Item (259/260/296/85-15-02)
This item will be tracked
under
which discusses
modifying valves
71-32
and
73-24
to
include
a testable
This unresolved
item is closed.
(Closed)
Unresolved
Item (259/260/296/85-15-03)
The licensee'
evaluation
and associated
calculations
were
reviewed
concerning
an
unmonitored
stack
release
for two hours.
No change
in release
rate
was
noted
during this
time.
This item is closed.
(Closed)
Violation (259/85-06-12)
Plant
procedures
were
reviewed
and
3.2.2,
Motor-Operated
Valves
Cycled
During
Cold
Shutdown,
and
MMI-51,
Maintenance
of CSSC/Non-CSSC
Valves
and Flanges
have
been revised to include
steps
to
remove
test
hoses
after maintenance
and testing.
This item is
closed.
(Closed)
Violation (259/260/296/85-06-08)
The licensee's
response
to this
violation was
reviewed.
Procedure
revision to Surveillance Instruction,
4.5.E. l.d and
e,
were
reviewed for clarification of the appropriate
times
for testing
the
High Pressure
Coolant Injection System
using reactor
steam
or auxiliary boiler steam.
This item is closed.
(Closed)
Unresolved
Item
(259/260/296/85-57-02)
The
question
concerning
divisional
separation
of electrical
cables
was resolved
by the licensee.
Section 8.A.4. 1 of the
FSAR covers
th'e special
case for cables
not leaving"
the
control
bay
that
have
adequate
fault protection
to
prevent
the
propagation of the fault.
This item is closed.
(Closed)
Inspector
Followup
Item
(259/260/296/85-57-12)
Health
Physics
Technicians
have
been
retrained
in the
methods
to be followed to document
and to determine
dose
assessment
of contamination.
A copy of the training
material
and
training
session
attendance
sheets
were
provided
to
the
inspector for review.
This fulfills the commitment
and this item is closed.
4.
Unresolved
Items" (92701)
There
is
an
unresolved
item
in paragraph
5.a.
concerning
updates.
Paragraph
9 contains
an
unresolved
item regarding
the trip point basis
of
radiation monitors.
"An Unresolved
Item is
a matter
about
which
more
information is required
to
determine
whether it is acceptable
or may involve
a violation or deviation.
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
wer'e held 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
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 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 Unit 3 Residual
Heat
Removal
System.
A complete
walkdown of the accessible
portions of the
Reactor
Protection
System
Panels 9-3, 9-4 and 9-5 was
conducted
to verify system operability.
Typical of the
items
checked
during the
walkdown were:
lineup procedures
match plant drawings
and the as-built configuration,
hangars
and
supports
housekeeping
adequate,
electrical
panel
interior
conditions,
calibration
dates
appropriate,
system
.instrumentation
on-line,
valve
position
alignment
correct,
valves
locked
as
appropriate
and
system
indicators functioning properly.
a.
Annual
FSAR Updates
During
a
review
of the
FSAR,
the
inspectors
noted that
a
change
submitted
as part of Amendment
1 deleted
an original
FSAR commitment to
periodically
perform
a
visual
inspection
of
secondary
containment
relief panels.
A further
review of Amendment
1 revisions 'detected
several
other similar examples.
These
are
noted below:
(1)
Section
5.3.5.1 of the
FSAR stated that the secondary
containment
relief panels
are visually inspected
periodically to ensure that
the panel s have not parti al ly rel i eved
and thereby
opened
cr acks
in the siding.
Amendment
1 changed this statement
to "The relief
panels
~ma
be visually inspected....".
(2)
Section
3'.5 of the
FSAR stated
that the
gas
pressure
in the
Standby
Liquid Control
(SLC)
System
is
measured
periodically to detect
leakage.
This
bladder-type
pneumatic-
hydraulic
is installed
near
each relief valve to
dampen
pulsations
from the
pumps to protect
the
system.
Amend-
ment
1
changed
this
statement
to
"The
gas
pressure
in
the
can
be measured....".
(3)
Section
5.3.5.3 of the
FSAR stated that the cooling water supply
to the equipment
area cooling units was initially tested with EECW
and
is
now tested
periodically
in
the
same
manner.
These
equipment
area cooling units
remove heat generated
by the
RHR and
core
spray
pumps to maintain the air at
les's
than
148 degrees
F.
Amendment
1 changed this statement
to "the cooling water supply...
can
be tested...".
(4)
Section 7.5.4.2.5 of the
FSAR stated that
a feature
which provides
for
a reactor trip signal"-to
be generated
by the
Source
Range-
Monitors (SRMs) is used for the
performance
of core alterations.
Amendment
1
changed
this
statement
to "this feature
~ma
be
used
during core alterations"
~
No
documented
justification
or
safety. evaluation
for these
changes
could be located
by the licensee.
Licensee
representatives
stated that
recently
developed
administrative
controls for'nnual
updates
(Standard
Practice
1. 13,
Final
Safety Analysis
Report
and Technical
Specifications)
require
detailed
evaluation
and justification for
changes.
This is considered
an unresolved
item (259/260/296/86-06-01)
pending proper evaluation of all previous
changes
to the
FSAR.
b.
Fire Hazard
Concerns
During
a routine
tour of the Unit 3 diesel
generator
building
on
January
8,
1986,
the
inspector
found several
cigarette
butts
on the
floor of the
3EA diesel
generator
room.
"NO SMOKING" signs
are clearly
visible
to
personnel
in
the
area,
and
plant
procedures
(Standard
Practice
14.56)
designate
this
area
as
a
non-smoking
area.
The
inspector's
concern
was
discussed
with the shift engineer.
Plant
management
later stated this item was discussed
in the shift turnover
meetings
and
anyone
caught
smoking in the
areas
would
be treated
as
deliberately violating plant procedures.
On January
14,
1986,
a fire occurred in
a 4160 volt line in the turbine
building.
The fault was cleared
by differential relay action
when the
breaker
feeding
shutdown
bus
one tripped.
The momentary disruption
in
voltage
caused
the
1A and
1B diesel
generators
to start but they were
not required to assume
the load.
The licensee is evaluating
the
cable
fault.
On January
17,
1986,
a fire occurred
in the Unit 2 reactor building,
elevation
519.
The fire
was
caused
by welding
cables
which
were
shorted
out
and arcing'he
cables
were
deenergized
and
the fire
extinguished.
The
problem
was
found to
be
an
unattended
energized:.
welding lead routed
so that the stinger
was
on
a
120 volt lighting cord
which was laying in
a pool of water.
As noted
by the licensee,
"The
general
condition of the, work .area
was in poor condition with several
obvious fire hazards
present
such
as frayed extension
cords,
extension
cords in pools of water,
unattended
and energized
welding equipment."
Areas
such
as these will be the focus of routine plant tours to insure
the licensee
is correcting potential fire hazards.
Incorrect Design Specification
Reference
On October 22,
1985, during
a routine review of plant drawing 730E927,
Isolation, -a- copy of a document
referenced
on the-
drawing was requested
from the,licensee
but could not be located.
The
document
was
a
design
specification,
22A1421,
for the
separation,
isolation,
and identification of engineered
safeguards.
The licensee's
search
for
the
document
and
discussions
with
General
Electric
representatives
revealed
that
design
specification
22A1421
was
not
applicable
to
Browns
Ferry.
The correct specification
was
22A2809.
Four design drawings were
found to reference
the incorrect specifica-
tion and are listed below:
730E918
Engineered
Safeguards
730E915
Reactor Protection
System
730E930
System
730E927
Primary Containment Isolation
The initial review of the problem by design
services
found significant
differences
between
the
two specifications
and possible
problems with
the correct
specifications.
The office of engineering
was
asked
to
review the problems but only performed
a limited review which revealed
no
significant
design
problem.
General
Electric
is
performing
a
thorough design analysis of the error for reportabi lity and operations
impact.
Design services
is tracking the resolution of the problem.
The failure to have adequate
design control to ensure that drawings
and
design
specifications
are
correct
was
noted
as
a
violation
of
Criterion III related
to
design
control.
(259/260/296/86-05-02)
d.
Ventilation Towers Design Error
The licensee
reported
on January
17,
1986, that
an engineering
analysis
disclosed
that
the
control
bay ventilation towers
are
not protected
from the effects
of tornado
generated
missiles.
The original
vent
tower design apparently did not include tornado missile protection for
heating,
ventilating,
and air-conditioning
(HVAC) equipment
in
the
towers.
Under cer tain conditions, thi s equipment could be required for
safe
shutdown.
FSAR Section
10. 12.3
requires
that
the
HVAC system
maintain
the control
bay
and electrical
board
rooms within acceptable
limits and allow occupancy
under all conditions.
The vent towers
open
and close
to act
as
makeup air ducts to
supplement
the
HVAC system.
The
licensee
is evaluating
the situation
as
to
the
severity
and
corrective action.
Inspector
Followup Item (259/260/296/86-05-10).
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
~--
accomplished
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.
Core Spray Testable
Check Valve Maintenance
The inspector
observed
Maintenance
Request
(MR) A-579826,
Disassemble,
clean,
inspect
and repair
as
necessary
the
Core
Spray Testable
Check
Valve (2-FCV-75-26) actuating cylinder.
No manufacturer's
drawings or
maintenance
instructions
could
be located for the actuator.
Mainte-
nance
personnel
contacted
the
manufacturer
and
learned
that
the
cylinder is
no
longer
made
and
were
unable
to obtain
drawings
or
instructions.
Maintenance
instructions
were
prepared
and
reviewed
by
Plant
Operations
Review
Committee
(PORC)
and
approved
prior to the
maintenance.
No
problems
were
found
during
the
work,
however
Mechanical
Maintenance will be evaluating
the
need for upgrading
the
actuators
since repair parts
are
no longer available.
b.
Rod Consolidation
Program
A
c.
Repair to 4160 Volt Unit Board Cable
d.
Unit 2 Outage
Access
Area Cleanup
No violations or deviations
were noted in this section.
\\
7.
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 for operation
met,
testing
accomplished
by qualified
personnel,
and
that
the
surveillance
was
completed
at
the
required
frequency.
SI 4.2.A-10B - Reactor Building Ventilation Radiation
Monitor Calibration
The inspector
witnessed
the performance
of SI 4.2.A-10B on the Unit 3
radiation monitors started
on January
22,
1986.
Numerous
problems
and
delays
were encountered
by the technicians
who performed
the
survei 1-
lance
including an inadvertent trip of the containment isolation logic
on January
23,
1986.
The Assistant Shift Engineer
(ASE)
stopped
the
surveillance
at
10:35
on
January
23,
1986
pending interpretation
of
procedural
inadequacies.
The
technician
and
instrument
maintenance
engineers
performed
a step-by-step
review of the procedure
and reached
a
consensus
on the interpretation
of -several
vague
instructions.
A
non-intent
change
to the
procedure
was initiated
and
the test
was
resumed
on January
24,
1986.
The calibration
was
completed
with
no
further incident.
Refer to paragraph
9, Calibration,
for
a discussion
of the procedural
deficiencies.
During the
inadvertent trip of the
containment
isolation
logic the
Standby
Gas
Treatment
System
Train
A and
the Control
Room
Emergency
Pressurization
-Train
A failed to initiate.
This
problem
is
being
investigated
by the licensee
and will be left as
an Inspector
Followup
Item (259/260/296/86-05-03)
pending resolution.
No violations or deviations
were noted in this section.
8.
Reportable
Occurrences
(90712,
92700)
The below listed licensee
events
reports
(LERs) were reviewed to determine
if the
information
provided
m'et
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
problems,-
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.
296/85-21
- 260/85-17
"260/85-14
260/85-03
259/85-44
t
259/85-23
"259/85-22
"259/83-59
Date
7/30/85
11/20/85
10/24/85
2/27/85
8/19/85
6/13/85
6/10/85
10/15/83
Event
Containment Isolation
because
of a fuse
removal
Engineered
Safety Feature
Actuations
from High
Radiation Alarms
Inadvertent
Containment
Isolation
Unplanned Initiation of
Reactor Protection
System
During Shutdown
Use of Unspecified Material
on Axial Restraints
Containment Isolation
Due
To Loss of Relay Neutral
Containment Isolation
(Comments
concerning this
LER are addressed
in IE
Report 85-49)
Inoperability of a Containment
Air Monitor
- 259/82-99
- 259/84-25
- 259/85-01
12/08/82
6/15/84
1/21/85
Leak found in a Cracked
1/2 inch Core Spray
Sample
.Connection
Pipe.
Inadequate
Isolation of
Building Heat System
Between
Reactor
and Turbine Building
in the
Drywell
(The details of this event
are
more accurately
described
in IE Report 85-06).
10
259/85-11
4/11/85
Primary Containment Isolation
System Initiation
The
inspector
reviewed
Licensee
Event
Report
Inadequate
emergency
equipment cooling water flow to the residual
heat
removal
and core
spray
coolers,
dated
October 4,
1985,
and
several
concerns
were
noted.
During
performance
of
a test
on
September
4,
1985,
to verify adequate
cooling water flow to components
supplied
by the emergency
equipment cooling
water
(EECW) system,
inadequate
flow was found to two Unit
(CS)
system
room coolers
and one Unit 3 residual
heat
removal
(RHR)
system
room
cooler.
The
room coolers
are required
by plant technical
specifications
to
'nsure
adequate
cooling to
RHR and
pump motors.
The
LER stated
that
an
evaluation
of past
data
would
be
performed,
tags
would
be
hung
on all
throttle valves that regulate
EECW flows to the
components,
and
a revision
would be
made to the test
procedure.
The inspector
reviewed the procedure
and found no revision.
Discussions
with cognizant
personnel
found that the
analysis
was not completed
and the tags
had not been
made.
The
LER gave
no
completion
date for these
items
and
cognizant
personnel
stated
that
the
completion
was believed
linked to restart
of Unit 2.
The inspector
could
find no, correlation
between
completion of the
items prior to restart
of
Unit 2 as
the specific
items in the
LER were applicable to other units
and
EECW is
a
common
system.
Discussions
with the licensee
contact
stated
on
the
LER found
no data
was given
on the licensee
commitmen't tracking system-
for completion of these
items.
The test
procedure
was
being
performed
every six weeks
and
the
problems
identified
on September
4,
1985.
The procedure
had been
performed
again
on
ll/2/85 and 1/7/86 although
none of the corwective actions
stated
in the
LER
were
complete.
This
LER will remain
open until all corrective actions
are
completed.
No violations or deviations
were noted in this section.
Calibration (56700)
The objective of this supplemental
module inspection is to ascertain
whether
the licensee
has developed
and implemented
a program for the calibration of
plant
instrumentation
that is in
conformance
with license
requirements,
technical
specifications,
license
commitments
and
industry
guides
and
standards.
This is confirmed by verifying that calibration frequencies
have
been satisfied,
documentation
is proper
arid maintained
and =procedures
are
technically adequate.
A partial
completion
of this
inspection
was
performed
and
documented
in
Inspection
Report 85-53,
November,
1985 'he
licensee
has determined that
deficiencies
identified
in that
inspection
have
a
potential
for being
programmatic
and
has instituted
a
100% review of all technical
specification
surveillance
instructions to verify that surveillance
requirements
are being
11
satisfied.
An Inspector
Followup Item will be opened
(259/260/296/86-05-04)
to review completion of the licensee's
program.
The central
issue for this
followup will
be
whether
the
licensee's
instructions
are
technically
adequate
to satisfy
the intent of technical
specification
surveillance
requirements.
In addition,
the
number
and
type of deficiencies
identified
during
a
review of
SI 4.2.A-10,
Reactor
Building Ventilation
Radiation
Monitor Calibration,
(discussed
below) exemplify a different concern related
to
whether
procedures
are
sufficiently detailed
and
technicians
are
sufficiently trained to allow the procedures
to be followed as written.
An
Inspector
Followup
Item (259/260/296/86-05-05)
will be
opened
to observe
critical Surveillance
Instructions during the preparation for Unit 2 Startup
in order to resolve this concern.
a
~
Reactor
Zone and Refuel
Zone Radiation Monitor Calibration
T.S. Table 4.2.A, Surveillance
Requirements
and
Reactor
Building Isolation Instrumentation,
requires
that the Reactor
Building Ventilation
High Radiation-Reactor
Zone
and
Refueling
Zone
instrument channels
be functionally tested
every
month
and calibrated
every three
months.
Two inconsistencies
in the technical
specifica-
tions were identified related
to this requirement.
Note
14 to Table
4.2.A is indicated
as
being applicable
to this requirement,
however,
the intent of this note
cannot
be
determined.
Note
14
states
that
"Upscale trip is functionally-tested
during functional test
time as-
required
by
Section
4.7.B.l.a
and
4.7.C.l.c".
Section
4.7.B. l.a
relates
to the
maximum pressure
drop across
the Standby
Gas Treatment
System
(SGTS) filters
and
is
in
no
way related
to
the
radiation
monitors.
Similarly,
Section
4.7.C. l.c
no
longer
exists
in
the
technical
specifications
since it was wevised to 4.7.C. 1.a in February,
1981.
Section
4.7.C. l.a requires periodic demonstrations
of secondary
containment
integrity
and
is
also
not
related
to
the
radiation
monitors.
This will
be
tracked
as
an
Inspector
Followup
Item
(259/260/296/86-05-06)
pending resolution.
An additional
Technical
Specification
inconsistency
was
noted after
a
problem developed
during the performance of SI 4.2.A-10.
The question
arose
as
to
how long the radiation monitor instrument
channel
can
be
left inoperable
during the surveillance
before
the
channel
had to
be
tripped or declared
Note
11 to T.S.
Table 3.2.A states
that the
channel
may
be
for up to four hours for survei 1-
lance without placing the trip system
in the tripped condition.
Note
22 to T.S. Table 4.2.A states
that
one channel
may be administratively
bypassed
for
a period not to exceed
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for functional testing
and
calibration.
4.2.A-10B,
Reactor
Building
Ventilation
Radiation
Monitor
Calibration states
in the Limitation and Actions paragraph
that when
a
channel
i s failed in an
unsafe
condition,
the trip system
containing
the unsafe failure may be bypassed
for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to allow testing
of
the
other trip
system.
Instrument
Maintenance
( IM) personnel
interpreted this statement
to
mean that they
had
up to
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
to
12
resolve deficiencies
discovered
during surveillance.
This will be left
as
an
Inspector
Followup
Item
(259/260/296/86-05-07)
pending
clarification of the T.S.
and the Surveillance Instruction.
The inspector
observed
the
performance
of SI 4.2.A-10B
on the Unit 3
radiation
monitors
started
on
January
22,
1986.
Many
of
the
step-by-step
instructions
are
vague
and require interpretation
by the
technician
performing
the
work.
The correct
interpretation
depends
heavily upon the training and expertise
of the technicians.
In several
instances
an incorrect interpretation
was
made.
(These
are discussed
below.)
The locations
specified
in the
procedure
for
connection
of
measuring
and test equipment
could not be ascertained
without reference
to the
vendor
maintenance
manuals.
These
manuals
are
not listed
as
necessary
references
in the procedure
and caused
some delay in the work
as the technicians
searched
for the appropriate
manuals.
The measuring
points
are
not readily accessible
and require
extreme
caution
on the
part of the technicians
in order to avoid inadvertent
shorting between
eyelet terminals
on circuit boards.
Even
though the technicians
were
aware of this potential
and
had emphasized
this fact to the inspector,
a short
from the test
caused
an inadvertent
downscale trip on
January
23,
1986
(Refer
to
paragraph
7,
Surveillance
Testing
Observation
for
a
discussion
of
the
inadvertent
ESF actuation).
Licensee
representatives
stated
that
a
long
term
program
has
been
initiated
to
review
and
modify equipment
as
necessary
to provide -"
testing jacks
on similar equipment
to reduce
the
frequency
of these
types of challenges
to safety
systems.
The
inspector
observed
the
following deficiencies
in the
way the
surveillance instruction
was written and performed:
(1)
Step
4. 1, Calibration
of the
Power
Supply The technicians
initially performed this section
on the
spare
power supply which
is listed in the
procedure
as
equipment
to
be
used
during
the
bench
test
of the
Sensor
and
Converter
Unit.
This
procedure
calibrates
the
+24 volts,
-24 volts,
and
+575
volts
power
supplies.
Although the
procedure
is unclear
as to whether this
section
applies
to the
spare
power
supply
used
during the
bench
test
or the installed
power
supply
which is
in service,
the
technicians
realized
that
the intent
of
the
procedure
is
to
calibrate
the installed
equipment
and correctly re-performed
the
'step
on the installed equipment.
Following the adjustment
of the
high voltage
to
+575 volts
using
a voltmeter
connected
to
an
internal terminal, the technicians failed to notice that the front
panel
meter
of the
power
supply
was indicating only about
470
volts.
Although the
procedure
does
not explicitly require that
this
be
checked,
a parenthetical
statement
indicated
that
the
front panel
meter
should
also
indicate
+575 volts.
When
the
inspector
noted this
problem,
the technician
sprayed
the meter
with an aerosol
static
charge
eliminator
and the
meter
deflected
to the correct
value of +575 volts.
The technician
noted that
static charges
on meter faces is
a
common occurrence.
,
13
(2)
(3)
Step 4.2, Calibration of the Indicator
and Trip Unit Although
not specifically stated
in the procedure,
the technicians
removed
the Indicator and Trip Unit from service
and
performed this step
on the bench with the spare
power supply.
The technician
assumed
that this
was
the
intent of the
procedure
since
the list of
required test equipment contains
the spare
power supply.
This was
later
determined
to
be
improper
since
the
upscale
and
downscale
trip setpoints
change
depending
on whether the Indicator and Trip
Unit is installed
in the
loop or bench
tested.
The technicians
determined
that
an upscale
setpoint
set at 90 mr/hr
on the bench
changes
to
about
60 mr/hr
when
installed
in
the
loop
and
a
downscale
setpoint
set
at ll mr/hr
on
the
bench
changes
to
off-scale
low when installed
in the
loop.
An additional
concern
r egarding
the calibration
of the trip setpoints
is
the
method
described
in the procedure.
The setpoints
are
adjusted
to about
10 mr/hr (downscale trip) and
90 mr/hr (upscale trip) using the
front panel
meter
on the Indicator and Trip Unit rather than using
calibrated
measuring
and test
equipment.
This front meter panel
is not calibrated
as part of the
surveillance.
The
instrument
calibration with
a radioactive
source
which is performed in Step
4.3.3 would provide
a check
on the meter;
however,
the tolerance
on this test is extremely broad
(+100%
50io).
Resolution of this
concern will depend
upon
the acceptability of the trip setpoints
(discussed
below).
Additi5nal caution
must be
u'sed
when using the
front panel
meter for adjustment of the trip setpoints
to avoid
an
erroneous
setting
due
to the static
charge
problem
discussed
previously.
The generic
aspects
of this problem will be tracked
as
an Inspector
Followup Item (259/260/296/86-05-08).
Step
4.3.2,
Electronic Calibration
of the
Sensor
and Converter
Unit Step b. of this section is worded awkwardly and allows one
to
skip
to
Step
e.
unless
the
procedure
is
being
performed
following maintenance
or installation
of
new
components.
The
technicians
misinterpreted
this
and
performed
Steps
c.
and
d.
anyway.
Step
e. of this
section
requires
the
equipment
to
be
bench tested with the equipment
connected
as
shown in Figure 4.3.
Although the
step
requires
a resistor
to
be
adjusted
until
a
positive deflection is observed. on the voltmeter,
no voltmeter is
shown
on the test setup,
Figure 4.3.
The technicians
interpreted
the voltmeter to
mean
the front panel
meter
on the indicator and
trip unit.
This is probably not the correct interpretation
since
the test
setup
shown in the vendor
manual
from which this part of
the
surveillance
was
derived
explicitly
shows
a
voltmeter
connected
to the recorder
output jacks
on the indicator and trip
unit.
Step f. of this section
states
to vary
the
voltage
and
observe
that
the indicator
and trip unit
goes
upscale
as
the
setpoint is reached.
When this step
was performed,
the indication
did not
go fully upscale
but
was erratic
and
tended
to
hang
up
14
around midscale.
An apparent
typographical error from a previous
revision left two steps
labeled
as f. in this section.
The intent
of the
second
Step f. could not be determined
and
was subsequently
determined
to be useless.
(4)
Step 4.3.3,
Source
Calibration -- The tolerance for this section
is
+7 '% equivalent linear full scale.
Due to the
logarithmic
nature of the panel
meter (4 or
5 decades
depending
on the model)
this translates
into
an
extremely
wide tolerance
band
(+100%,
-50%)
on
the
indicated
dose
rate.
The Technical
Specification
upscale trip setpoint for these
radiation monitors is less
than
100 mr/hr.
The Standard
Industry practice for selecting
setpoints
is to select
a setpoint which is less
than
the required
setpoint
by
an
amount at least
eqaivalent
to the accuracy
of the device.
In this case,
that would be
50 mr/hr..
The licensee
has selected
a
setpoint
of
92
mr/hr.
The
basis
for this
setpoint
is
not
documented.
This will be left as
an Unresolved
Item (259/260/296/
86-05-09)
pending
a review of the basis for the setpoint
when this
information is developed
by the licensee.
The
above
items were discussed
in detail with licensee
representatives
who indicated that procedures
for all radiation monitors are generally
weak
as is technician
expertise
in this area.
These
procedures
have
been .assigned
a
high priority in the
procedures
upgrade
program;
-'owever,
revisions
are not expected
to be completed
immediately.
b.
Review of Previous
SI 4.2.A-10 Data
The inspector
reviewed data
sheets
for- the last five completions of SI
4. 2.A-10,
Reactor
Building Ventilation Radiation
Monitor, Functional
Test.
The following inconsistencies
on the recorded
data
were noted:
(1)
I
On about half of the survei llances
reviewed the
bug source
values
recorded
in Step
4. 1.C were in error
by
a factor of ten for the
RM-90-142 and
143 radiation monitors.
The
bug
source
is
a
small
radioactive
source installed in. the radiation monitor to maintain
the indication slightly upscale
so that
an instrument failui e
can
be
detected
by
a downscale
condition.
The
bug
source
value is
recorded
so that it can
be added to the calibrated
source
used in
the
procedure
to determine
the actual
radiation
exposure
to the
detector
tube.
As
an
example
of
the
error,
the
bug
source
recorded
for RM-90-142
on
9/3/85
was
5 mr/hr.
Later in the
surveillance
in
Step
4.2.a,
the
meter
indication
was
again
recorded
and
compared with the stripchart recorder
as
a check
on
the recorder calibration.
This data
was recorded
as 0.5 mr/hr,
a
factor of 10 less
than previously recorded.
A review of the data
using
correct
bug
source
values
indicated
that
the
instrument
calibration
was still within tolerance
on those occasions.
15
(2)
In Step 4.2.a
and
b the indicator
and trip unit is recorded
and
the indication
on the stripchart
recorder is recorded.
Although
the intent of these
steps
is to compare
the
two readings
as
a
check
on the recorder calibration, this is not explicitly stated
and
no criteria is
given for
how closely
they
should
match.
Inconsistencies
in the
recorded
data
shows that
sometimes
this
data
is
the
bug
source
value
whereas
other
times it is
much
greater.
This is due to the step
immediately preceding
Step 4.2.a
in which the radiation monitor is exposed
to
a check source.
The
procedure
does
not state
when to remove the check source
so
some
technicians
do it before this step
and others
do it afterwards.
(3)
The
data
sheet
on
the
Unit
1 radiation
monitors
performed
on
11/4/85
recorded
the
upscale trip setpoints
at
100 mr/hr.
The
Surveillance
Instruction requires
the setpoint to be less
than
92
mr/hr .
Subsequent
performances
of the
surveillance
have
since
reset
the trip setpoints
to less
than
92 mr/hr.
The fact that these recurring errors were not detected
by reviews
of
the
Instrument
Mechanic
Foreman
and
Cognizant
Engineer
indicates
a
lack of attention
to detail
in
these
reviews.
Although none of these errors resulted
in an adverse
impact
on the
equipment
calibration,
this is considered
fortuitous
since
the
potential
was there.