ML17219A508
| ML17219A508 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 03/24/1987 |
| From: | Bassett C, Hosey C, Revsin B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17219A506 | List: |
| References | |
| 50-335-87-04, 50-335-87-4, 50-389-87-04, 50-389-87-4, NUDOCS 8704130278 | |
| Download: ML17219A508 (20) | |
See also: IR 05000335/1987004
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.IlII.
ATLANTA,GEORGIA 30323
r, r.
)Qt'~s
Report Hos.:
50-335/87-04
and 50-389/87-04
Licensee:
Florida Power and Light Company
9250 West Flager Street
Miami, FL
33102
Docket Nos.:
50-335
and 50-389
Facility Name:
St. Lucie
1 and
2
License Nos.:
and NPF-16
Inspection
Conducted:
March 2-6,
1987
Inspectors:
2&, I
j; C..
as t~
(.
Luau~
evs>n
/:
Approved by:
C.
M. Hosey,
Sects
n
C ief
Division of Radiat'on Safety
and Safeguards
pity,/J
Date Signed
8/o
/8".
ate Signed
SUMMARY
Scope:
This was
a routine,
unannounced,
radiation protection inspection in the
areas
of external
exposure
control; internal
exposure
control;
control of
radioactive
material,
contamination
surveys
and
monitoring;
program for
maintaining exposure
as
low as reasonably
achievable
(ALARA); solid radioactive
waste
handling
and disposal;
transportation
of licensed material;
inspector
followup items;
IE Bulletins, Notices,
and allegation followup.
Results:
Three violations
were identified:
failure to label
containers
of
radioactive
material,
failure
to
perform
surveys
per
Department
of
Transportation
(DOT) requirements,
and failure to implement
an adequate
guality
Control
(gC) program for waste characterization
and to properly solidify waste.
8704130278
870327
ADOCK 05000335
G
REPORT DETAILS
Persons
Contacted
Licensee
Employees
- K.
- D
- R.
- l
- C
- p
- H
- L
- J
- R
- R.
- E
- C
+J
- S
- C
- R
- M
- A;
- p
- H.
- L.
L.
K.
L.
J.
- L
D.
D.
D.
N. Harris, Site Vice President
A. Sager,
Plant Manager
Sipas,
Services
Manager
A. Dillard, Maintenance
Superintendent
L. Wilson, Assistant Maintenance
Superintendent
D. Parks, Backfit Manager
F. Buchanan,
Health Physics Supervisor
W. Pearce,
Operations
Supervisor
Scarola,
Electrical Maintenance
Supervisor
J. Frenchette,
Chemistry Supervisor
A. Symes,
equality Assurance
Supervisor
J. Wunderlich, Reactor Engineering Supervisor
A. Pell, Technical
Supervisor
Krumins, Site Engineering Supervisor
C. Sanders,
Mechanical
Maintenance
R. Siebold, guality Assurance
Engineering
Dawson, Electrical Maintenance
Synder,
Engineer
J.
Gould, Corporate Health Physics
J. Stoner,
Corporate
Health Physics
M. Mercer, Health Physics
Technical
Supervisor
L. Large, Health Physics Assistant Operations
Supervisor
R. Baker, Health Physics Administrative Supervisor
W. Payne,
Health Physics
ALARA Technician
E.
Pugh,
Health Physics'nstrument
Supervisor
R. Smith, Health Physics
Radiation Protection Supervisor
E. Jacobus,
Health Physics
ALARA Technician
West, Shift Technical
Advisor Group Lead Engineer
Haithcox, Health Physics
Radioactive
Waste Technician
Spaugh, guality Control
A. Bailey, guality Assurance
B.'arks, Quality Assurance
Other
licensee
employees
contacted
included,
technicians,
operators,
security force members,
and office personnel.
S. Nuclear Regulatory
Commission
- H. Bibb, Resident
Inspector
- Attended exit interview
3.
-
4 ~
Exit Interview
The inspection
scope
and findings were
summarized
on'March 6,
1987, with
those
persons
indicated
in
Paragraph
1
above.
Violations involving
failure to
label
B-25
metal
boxes
containing
radioactive
material
(Paragraph
6), failure to comply with
DOT requirements
for surveying the
undersides
of transport
vehicles
(Paragraph
9),
and failure to properly
solidify waste
and
to
implement
an
adequate
gC
program for waste
characterization
(Paragraph
8) were discussed
in detail.
The quality of
written and
approved
radiation protection
procedures,
and
the
use of
internal
administrative
guidelines
and
memoranda
were
discussed
with
management.
The licensee
committed to review and revise
as necessary
the
health
physics
procedures
and to send the
NRC schedules
for completion of
the review and for procedure
implementation the first week of April 1987.
It was anticipated
that implementation
would
be
complete
by the
end of
1987.
The licensee
acknowledged
the inspection findings'nd
took
no
exceptions
to the apparent
violations except for the violation concerning
dose
rates
on the undersides
of vehicles stating that surveys of these
areas
were not necessary
to know that dose
rates
were within the limits.
The licensee
did not identify as proprietary any of the materials
provided
to or reviewed
by the inspector during this inspection.
Licensee Action on Previous
Enforcement Matters
(Closed)
Violation (50-335/86-01-01)
Failure
to
Perform
an
Adequate
Evaluation of Personnel
Whole
Body Exposure.
The inspector
reviewed the
licensee's
response
dated
May 27,
1986,
and verified that the corrective
actions in the response
had been
implemented.
(Closed)
Violation
(50-335/86-09-01
and
50-389/86-08-01)
Failure
to
Maintain Written Procedures
for Respiratory
Protective
Equipment
Issuance
Records.
The inspector
reviewed
the licensee's
response
dated July 30,
1986,
and verified that the corrective action indicated in the
response
had been
implemented.
(Closed) Violation (50-335/86-09-03
and 50-389/86-08-03)
Dose
Rates
on the
External
Surface of Packages
of Radioactive Material Offered to
a Carrier
for Transport
in
Excess
of
DOT Limits.
The inspector
reviewed
the
licensee's
response
dated July 30,
1986,
and verified that the corrective
action specified in the response
had
been
implemented.
(Closed)
Violation
(50-335/86-09-04
and
50-389/86-08-04)
Failure
to
Package
LSA Radioactive Material in a Strong Tight Package.
The inspector
reviewed
the licensee's
response
dated July 30,
1986,
and verified that
the corrective action specified in the response
had been
implemented.
External
Exposure Control
and Dosimetry (83724)
a.
specifies
the applicable
radiation
dose
standard
for
individuals in restricted
areas.
The inspector
reviewed the computer
printouts
(Form
NRC-5 equivalent)
for the current
calendar
year,
1987,
and verified that the radiation
doses
recorded for plant
and
contractor
personnel
were within the quarterly limits of 20.101(a).
Selected
Form
NRC-4s were also
reviewed
and it was determined that
exposure histories
were being completed
and maintained
as required
by
10 CFR 20.202 requires
each
licensee
to supply appropriate
personnel
monitoring equipment to specific individuals
and to require the
use
of such
equipment.
During plant tours,
the
inspector
observed
workers wearing thermoluminescent
dosimeters
(TLDs) and self-reading
dosimeters
(SRPDs)
as required.
For maintenance
activities
involving steam
generator
work, the health
physics
procedure
HP-7,
Health
Physics
Requirements
for All Steam
Generator Activities,
Rev. 0, dated April 4, 1986, required the
HP Supervisor to determine
the
number
and location of TLDs and dosimeters
used
by workers
based
upon
an initial survey of the area.
The dosimetry requirements for
sludge
lancing operations
were more specific in Procedure
HP-7 with
specific
locations
identified for
multi-badging
'and
extremity
dosimetry.
HP coverage
of such operations
was also specified.
All
entries into the secondary
handholes
required constant
HP coverage
by
a qualified Senior Health
Physics
Technician.
While observing
the
steam
generator
work and
sludge
lancing operations
in Unit 1, the
inspector
verified that
the
workers
were
wearing
the
required
dosimetry in the locations specified
and that continuous
HP coverage
was being provided.
requires
that
each
licensee
make or cause
to
be
made,
such
surveys
as
may
be necessary
for the licensee
to comply
with the regulations
and are
reasonable
under the circumstances
to
evaluate
the extent of the radiation hazards
that
may be present.
A
survey
is
defined
in
as
an evaluation
of the
radiation
hazards
incident to the production,
use,
release,
disposal
or presence
of radioactive materials
or other sources
of radiation
under
a specific set of conditions.
10 CFR 20. 101(a)
requires
that
no licensee
possess,
use or transfer
licensed
material
in such
a manner
as to cause
any individual in a
restricted
area to receive in any period of one calendar quarter from
radioactive material,
a total occupational
dose in excess of 7.5
to the skin of the whole Lody.
The inspector
reviewed the licensee's
procedure for calculating
dose
to the skin, HP-72, Determination of Dose to the Skin From Fixed Skin
Contamination,
Revision 2,
dated
March 2,
1987.
It was noted that
the procedure
had
been
revised
to require calculation of dose to the
skin from fixed contamination
in excess
of 10,000
counts
per minute
per
probe
area
and
from contamination
of the skin
due
to
a hot
particle.
A hot particle
was
defined
as
a very small
piece of
radioactive
material
that
had
high radiation levels that extended
over short distances
and it was noted that, in many instances,
the
particle would not
be visible to the
naked
eye.
The procedure
did
not require skin dose to be determined if the contamination
detected
could
be
removed i.e., did not remain fixed in the skin.
During
discussions
with the
licensee,
the
inspector
learned
that
no
threshold
levels
had
been
established
to require
an
assessment
of
skin
dose
due to removable
contamination
and that
such
assessments
were not routinely performed at the facility.
The inspector
reviewed
selected
licensee
and contractor
personnel
skin/clothing contamination
reports for calendar year 1987.
It was
noted that
on February 20,
1987,
a licensee
employee
had detected-
contamination
on the left side of his face
as
he was frisking out of
the radiation control
area.
The individual
had
been in the Unit
1
Reactor
Containment
Building (RCB) for approximately three
hours
and
twenty
minutes
and
had
been
wearing
protect'ive
clothing
which
consisted of rubber shoe covers, coveralls,
gloves
and
a cap, instead
of
a
hood.
The worker apparently
became
contaminated
whil'e laying
prone at the wall of the reactor cavity to perform an inspection.
The worker was taken to the personnel
decontamination
area
where the
initial level of contamination
was
determined
to
be
one million
disintegrations
per minute
(dpm).
Nasal
smears
were taken
but
no
contamination
was detected.
The contaminated
area
was subsequently
decontaminated
and
a re-survey of the area indicated
no contamination
remained.
A whole
body
count
was
conducted
with no detectable
internal
deposition of radioactivity.
An incident report
was also
completed
but no skin dose calculation
was performed.
The licensee
determined
that
the
contamination
was
not attributable
to
a hot
particle but was uniformly distributed over a portion of the hair and
skin of the left cheek.
The inspector
determined that,
assuming
twenty square
centimeters
as
the contaminated
area
(the approximate
area of a probe), cobalt-60
as the isotope
involved and three
hours
and twenty minutes
as
the length of time the contamination
remained
on the
skin (worst case),
the
dose
to the
skin would
have
been
approximately
750 millirem.
The inspector
discussed
the incident with licensee
representatives
who indicated
that
the
subject
of skin
dose
assessment
due
to
removable
contamination
was
being
reviewed.
The
licensee
had
discussed
the matter with members
of the Corporate
Health
Physics
Staff and
was in the process of determining
an appropriate
threshold
level for requiring
skin
dose
assessments
which would then
be
incorporated
into
a procedure.
The licensee
also indicated that
previous
instances
of skin contamination
would
be
evaluated
to
determine
whether any regulatory limit had been
exceeded.
The licensee
was
informed that failure to assess
skin
dose
from
contamination
would
normally
be
considered
a violation of the
requirements
of
However,
the
NRC Enforcement
Policy delineated
in 10 CFR 2, Appendix C, 1986, states
that
a Notice
of Violation will generally
not
be issued for violations identified
by the licensee
provided that the licensee identification meets
the
criteria
specified
by
The inspector
stated
that this
apparent
violation met the required criteria
and consequently
would
be considered
licensee identified.
The licensee's
corrective action
will be reviewed during future inspections
(50-335, 389/87-04-01).
5.
Internal
Exposure Control
and Assessment
(83725)
a ~
b.
10 CFR 20. 103(a) establishes
the limits for exposure
of individuals
to concentrations
of radioactive
materials
in air in restricted
areas.
This section
also
requires
that suitable
measurements
of
concentrations
of radioactive materials in air be performed to detect
and evaluate
the airborne radioactivity in restricted
areas
and that
appropriate
bioassays
be
performed to detect
and
assess
individual
intakes of radioactivity.
The
inspector
reviewed
selected
results
of general
in-plant air
samples
taken
during
calendar
year
1987
and
the results
of air
samples
taken
to support Unit
1 steam
generator
work authorized
by
specific
radiation
work permits.
The
inspector
also
reviewed
selected
results of whole body counts
and the licensee's
assessment
of individual
intakes
of radioactive
material
performed
during
calendar year 1987.
10 CFR 20. 103(b)
requires
the
licensee
to
use
process
or other
engineering
controls,
to
the
extent
practicable,
to limit
concentrations
of radioactive
material
in air to levels
below that
specified
in Part 20,
Appendix B,
Table I,
Column
1,
or limit
concentrations,
when
averaged
over the
number of hours
in
a
week
during which individuals are in the area,
to less
than
25 percent of
the specified concentrations.
The
use
of* process
and
engineering
controls
to limit airborne
radioactivity concentrations
in the plant was discussed
with licensee
representatives
and the
use of such was observed
during tours of the
plant.
No violations or deviations
were identified.
6.
Control of Radioactive Materials
and Contamination
Surveys
and Monitoring
(83726)
a
~
20.401
and
20.403
require
the licensee
to perform
surveys
and to maintain records of such surveys
as necessary
to show
compliance with regulatory limits.
The Final Safety Analysis Report
(FSAR) of Units
1
and 2,
Chapter
12, outlines
survey
methods
and
instrumentation
while each Unit's Technical
Specifications
(TS) 6. 11
requires
adherence
to written procedures
for all operations
involving
personnel
radiation exposure.
During plant tours,
the
inspector
examined
radiation
levels
and
contamination
survey results
posted at the entrance
to the Unit
1
radiation
control
area
(RCA).
The inspector
also
reviewed
the
b.
c ~
results
of selected
surveys
taken in support of the
work in Unit 1.
Selected
Radiation
Work Permits
(RWPs) controlling
general,
as
well
as
specific radiological activities
were
also
reviewed.
The inspector
observed
the
use of survey instruments
by
plant staff and examined calibration stickers
on radiation protection
instruments
in use
by licensee
personnel.
Instrument
use appeared
to
be adequate
and all instruments
examined
had been calibrated.
The inspector
reviewed
the
procedure
which specified
the
release
criteria for items to be released
from an
RCA,
HP-41,
Movement of
Material
and
Equipment,
Revision 6, June 2,
1986.
While touring the
plant
and
surrounding
areas,
the inspector
observed
health physics
technicians
surveying
items
to
be
removed
from the
RCA.
Through
observation
of and
discussions
with various
technicians, it was
determined
that
adequate
release
surveys
were being
performed
and
that items with inaccessible
surfaces
were apparently
not released.
The inspector also
observed
workers exiting the
RCA from Unit l.
A
two minute frisk was required after leaving the contamination control
area
and another frisk of the
hands
and feet was required to leave
the
RCA.
The inspector discussed
the adequacy of the personal
survey
using
a frisker and hand-held
probe
due to the difficulty in frisking
the
back with the short-handled
probes
at the control point.
The
licensee
stated
that there
had
been
no problems to date but stated
that consideration
was being given to acquiring
a 'number of Eberline
personal
contamination monitors
(PCM-1s) for personnel
surveys.
specifies
the posting
and control
requirements
for
radiation
areas,
high radiation
areas
and airborne radioactivity
areas.
Additional requirements
for the control of high radiation
areas
are contained
in both units'S 6.12.
d.
During tours of the plant
and observation of work in Unit
1
RCB, the
inspector
reviewed
the licensee's
posting
and control of selected
radiation,
high radiation
and
airborne
radioactivity
areas
and
performed
independent
radiation
surveys
using
NRC equipment.
The
inspector's
measurements
agreed
with those
of the
licensee.
The
security of selected
was also
checked
and
found to meet the requirements
of TS 6. 12.
states
that,
except
as
provided
by 20.203(f)(3),
each
container of licensed
material
shall
bear
a durable,
clearly
visible label identifying the radioactive contents
and shall
bear the
radiation
caution
symbol
and
the
words
"Caution"
or
"Danger,
Radioactive Material,"
and shall
provide sufficient information to
permir. individjsls us, w. ir handling
the containers,
or working in
the vicinity thereof,
to take
precautions
to avoid or minimize
exposures.
exempts
labeling of containers
for containers
that
do not contain
licensed
material
in quantities
greater
than
7
applicable
limits specified
in
Appendix C,
and for
containers
when they are
in transport'nd
packaged
and labeled
in
accordance
with DOT regulations.
During tours of the plant
and the
RCB, containers
of radioactive
material
were checked for proper labeling.
In general,
containers
or
packages
were
labeled
as
required
except for two locations.
On
March 3,
1987,
28 B-25 metal
boxes
located
behind Unit 2 and
12 B-25
metal
boxes
located
adjacent
to
the
Steam
Generator
Blowdown
Treatment
Facility were
not labe1ed
as
required for radioactive
material
nor were
DOT labels applied.
At both locations,
the areas
were
barricaded
by ropes
bearing
the postings,
Radiation
Area
and
Radioactive Materials Area.
Through discussions
with the licensee it
was
determined
that the
boxes
were in areas
designated
as temporary
storage
areas
awaiting disposal.
Review of licensee
surveys of the
boxes
indicated
that radiation levels
up to 700 millirem per hour
(mr/hr)
on contact
had
been
detected
but
had
been
stacked
such that
these
dose
rates
were
inaccessible
to personnel.
The inspector
surveyed
selected
boxes
and noted
a radiation level of 100 mr/hr at
contact with one box.
Radia'tion levels
on the
boxes
indicated that
the radioactive contents
were in excess
of Appendix
C limits and that
the
exemptions
specified
in
were therefore
not
applicable.
The inspector
informed the licensee
that failure to
label containers of radioactive material
was
an apparent violation of
10 CFR 20.203(f) (50-335, 389/87-04-02).
7.
As Low As Reasonably
Achievable
(ALARA) (83728)
10 CFR 20. 1(c)
specifies
that
licensees
should
implement
programs
to
maintain worker's
dose
Other
recommended
elements
of an
program are contained in Regulatory
Guide 8.8 and 8. 10.
Chapter
12 of the
two Units'SARs also contain licensee
commitments
regarding worker ALARA
actions.
The inspector
reviewed Administrative
Procedure
No. 3300120,
St.
Lucie
Plant
ALARA Program,
Revision 4,
June 20,
1983,
which contained
the
elements of the
ALARA program.
The focus of the
ALARA program is through
the
ALARA Review Sheet,
a form that is required for all Radiation
Work
Permits,
and which mandates
the type
and depth of .ALARA review required
for the job.
The inspector
reviewed the minutes 'of the quarterly meetings of the
Review Board for 1986.
The Board is
composed of plant department
heads
and contractor project leaders
and considers
dose reduction
as it relates
to routine operation,
outage
planning
and facility design modifications.
The inspector
noted that
attendance
at
these
meetings
was
good.
The
licensee
stated
that the
Board also participates
in the
development
of
long range
ALARA plans
and
had adopted'an
exposure
reduction
program with
a goal of 285 man-rem per reactor
by 1990.
The
licensee
stated
that
the
plant
ALARA group
concentrat'ed
their
attention
on
outage
preplanning
and
had
made
major strides
in
dose
reduction for reactor
head
work through
improved shielding.
Additionally,
chemical
decontamination
of the
had been
planned for the
February,
1987, refueling
outage
which consisted
of flushing the system
with
NH OH, followed by hydrolasing.
Due to time constraints,
only steam
Generator
A was treated.
A dose rate reduction of 30 percent
was achieved
on the hot leg
and
35 percent
on the cold leg.
It was anticipated that
these efforts would be expanded for future outages.
The collective dose
measured for the site in 1986
(by TLD) was 469 man-rem
or 235 man-rem per reactor.
For 1987 the man-rem goal'is
884 (442 man-rem
per reactor).
The increase
in 1987
was
due to two outages
that have
been
scheduled
for the year,
one of which includes
a
10-year
In-service
Inspection.
No violations or deviations
were identified.
8.
Solid Waste
(84722)
a.
requires
any generating
licensee
who transfers
radioactive
waste to
a land disposal'acility to prepare all wastes
so that the waste is classified
according to
and meets
the waste characteristic
requirements
in 10 CFR 61.56.
requires
waste to have structural stability which
will generally
maintain its physical
dimensions
and
form under
expected
disposal
conditions.
states
that liquid wastes,
or wastes
containing
liquids, must
be converted into a form that contains
as -little free
standing
and noncorrosive liquid as is reasonably
achievable,
but in
no
case
shall
the liquid exceed
one
percent of the
volume of the
waste
when
the waste
is in
a disposal
container
designed
to ensure
stability, or 0.5
percent
of the
volume of the
waste for waste
processed
to
a stable form.
On
November 25,
1986,
the
licensee
shipped
two
metal
liners
containing
sludge
to
a
land
disposal
facility (Barnwell,
SC).
Radioactive
Waste
Shipment
No. 86-61
was specified
on the shipping
manifest
as
Radioactive
Material,
low specific activity
(LSA),
n.o.s.,
UN 2912, described
as
sludge solidified with cement,
Class
A
stable,
and
was transported
as
Exclusive
Use
on
a flatbed trailer.
Total
radioactivity
in
the
shipment
was
0.0180 curies.
Upon
inspection of the
two liners
by
an inspecto'r
from the State of SC
when
they arrived at the burial facility, it was
found that the
contents
had failed to solidify per the Process
Control
Program
(PCP)
as .evidenced
by
a paste-like
material
flowing from one liner when
punctured.
The second liner was suspect..
The burial site is =prohibited
by the State
of
SC from receiving
,unsolidified sludge
and consequently
the two liners were returned to
the licensee's facility.
On December 2, 1986, the State of SC issued
a violation to the licensee,
imposed
a civil penalty of one thousand
dollars
and prohibited further shipments
of solidified sludge to the
burial ground until acceptable
corrective action
had
been achieved
by
the licensee.
The suspension
of burial privileges for this waste
form was
rescinded
on January
13,
1987,
by the State of
SC after
review -of corrective measures
proposed
by the licensee.
When notified of the failure to solidify, the licensee
dispatched
a
representative
to the burial
ground to confirm the finding.
Upon
return of the two steel liners to the plant site, the liners were cut
open for inspection
of the contents.
, Inspection
showed that
one
liner failed to solidify at all (the liner punctured
by the State of
SC) while the
second
contained
a
mass
equal
to approximately
85
percent solidification.
Failure to insure
waste structural stability and failure to convert
the waste into
a form such that the waste containing liquid did not
exceed
0.5 percent of the
volume of the waste
was identified as
an
apparent violation of 10 CFR 20.311(d)
(50-335, 389/87-04-03).
requires
any generating
licensee
who transfers
radioactive
waste to
a land disposal facility to conduct
a quality
control
(QC)
program
to
ensure
compliance
with
and
61.56.
specifies
the
minimum
requirements
for
waste
characteristics
for all classes
of waste.
Upon return of the
two metal liners of Shipment
No. 86-61 to the
plant,
the Quality Assurance
(QA) Department
conducted
an audit to
determine
the
reasons
why satisfactory solidification had not been
achieved.
Solidifications
were
performed for the
licensee
by
a
vendor.
Audit No. QSL-OPS-86-491
specified
that
the
problem
was
two-fold:
mechanical
and chemical.
The mechanical
problem was that
solids from the sludge
stuck to the liner filters resulting in caking
in the bottom of the liner.
Consequently,
mixing of one liner was
incomplete
and
thus
the test
solidification
samples
were
not
representative.
The chemical
problem was that of ammonia.
The dried
sludge that
had
been
mixed with water in the liner had originated
from the
sewage
treatment
system
and consequently
had
a high ammonia
content.
The ammonia content altered
the
pH of the mixture such that
the exothermic
reaction
necessary
for solidification was inhibited.
The licensee's
QA audit
concluded
that:
(1) procedures
for the
vendor's
PCP lacked sufficient qualitative
and quantitative criteria
to assure
satisfactory
accomplishment
of the solidification process;
(2) there
were
no provisions for independent verification in Process
Control
procedures
to assure
that solidification was satisfactorily
10
accomplished;
and (3) quality records
were inadequate,
inaccurate
and
incomplete.
The inspector
reviewed Administrative Procedure
No. 0520025,
Process
Control
Program,
Revision 5, July 29,
1985.
Paragraph
8.2 states
that solidification, encapsulation
or absorption of radioactive waste
materials
shall
be
performed
in accordance
with vendor
approved
procedures.
The inspector
also
reviewed
Health
Physics
Procedure
No. HP-40,
Shipment
and Receipt of Radioactive Material, Revision 29,
July 25,
1986,
which stated
in
Paragraph
4.5 that containers
of
non-compactable
radioactive
waste
shall
be verified to be free of
standing water or oil by two individuals,
one of which will be Health
Physics
(HP).
The inspector also reviewe'd the vendor procedures
and
noted that there
was
no procedure specific for sludge solidification.
The licensee
stated that this had
been
noted at the time and that the
vendor operator
had contacted
his office and
was told to use
one of
the existing procedures
applicable to aqueous
wastes
since the sludge
was
water soluble.
This
procedure
specified
the
steps
for the
operator
to
follow in
performing
a
test
solidification
and
solidification of the final product, but did not include
gC checks to
be performed
by the licensee.
Review of calculation
sheets
used
by the
vendor operator
revealed
computational
errors for amounts
of chemical
additions
and
also
transcription errors
in transferring
numbers
from one worksheet
to
another.
Additionally, review of the operator's
log
book which
specified activities
actually
performed
by
the
operator
showed
inconsistencies
between it and
the calculation
sheets
and
was in
general, difficult to decipher.
The licensee
stated that their first
point of interaction with the vendor operator
required
by procedure
was at verification of solidification of the final product which was
accomplished
by visual
observation
and
by prodding
the top of the
product with
a stick.
Consequently,
vendor errors
had
not
been
detected.
The inspector stated that although the licensee's'CP
had
been
approved
by the
NRC, site specific procedures for implementation
of the
were
necessary
and should consider actions
necessary
to
insure
by
performance
and/or verification that 'appropriate
waste
stabilization
had
been
accomplished.
Areas
discussed
included
calculations,
representativeness
of test
samples,
solidification of
test
samples,
solidification of final product,
methodology
and
criteria to
be
used for pronouncing final product solid and vendor
procedures
appropriate
to the waste form being solidified.
Failure
to
conduct
a quality control
program
to
insure
waste
characterization
in accordance
with 10 CFR 61.56
was identified as
an
additional
example
of
an
apparent
violation of
(50-335, 389/87-04-03).
11
Transportation
(86721)
requires
each
licensee
who transports
licensed
material
outside of the confines of its plant or other place of use to comply with
the applicable
requirements
of the regulations
appropriate
to the
mode of
transport of DOT in 49. CFR Parts
170 through
189.
49
CFR 173.475(i)
states
that before
each
shipment of any radioactive
materials
package,
the shipper shall
ensure
by examination or appropriate
tests
that
external
radiation
and, contamination levels're
within
allowable limits.
49
CFR 173.411(b)(2)
specifies that radiation levels at any point on the
outer surface
of exclusive
use vehicles,
including top and underside
of
the vehicle,
must not exceed
200 millirem per hour during transportation.
h
The inspector
reviewed selected
records of radioactive materials
shipments
and radioactive
waste
shipments
made during September
through
December
1986.
It was
noted that the following shipment
records failed to denote
radiation
levels
taken
on
the
underneath
side
of
the
vehicle:
(1) No. 86-46
on
September
23,
1986;
(2) No. 86-47
on October 2,
1986;
(3) No. 86-49
on
October 9,
1986;
(4) No. 86-52
on
October 14,
1986;
(5) No. 86-53
on October 21,
1986;
and (6) No. 86-54 on October 23,
1986.
The licensee
stated
that they were
aware that surveys
on the bottoms of
trucks were required
by the regulations
but could not recall whether they
had
been
performed for the specific
shipments
in question.
The
Supervisor
stated
that the source of the problem lay in the survey forms
in that
no spaces
were clearly delineated for survey readings
taken
on the
bottom of transport vehicles.
Failure
to insure
by examination
or appropriate
tests
that external
radiation levels
on the
underneath
side of the transport
vehicle were
within the allowable limits was identified
as
an apparent
violation of
(50-335, 389/87-04-04).
Followup on IE Bulletins (92703)
(Closed)
BUL (50-335/78-19-08)
IE Bulletin 78-08 required
licensees
to perform
a review of shielding
design
of plant
areas
adjacent
to fuel transfer
tubes
to identify
potential
high radiation
areas,
both
continuous
and transient,
assure
positive control of the areas,
conduct
special
surveys
and
provide
a
written response
of the findings and actions
to resolve
any problems to
the
NRC.
The licensee
response
of August 11,
1978, indicated that (1) investigation
of the shield design
and the radiation associated
with the fuel transfer
tube would be completed prior to the April 1979 refueling and (2) the fuel
transfer
tube
area
would
be
surveyed
during the April 1979 refueling.
.
These
actions
were
completed
by the licensee;
however,
resurvey of the
12
fuel transfer tube area
was necessary
due to missing documentation for the
initia 1 survey.
The
i'nspector
reviewed
the
surveys
of the Unit 1 fuel transfer
tube
conducted
from April 14 to April 27,
1979,
documented
in the licensee's
Corrective
Action
Commitment
Request,
06-25-78,
June
22,
1978
and
referenced
in
Inspection
Report
50-335/79-3,
January 3,
1980.
The
radiation surveys
appeared
to be adequate
and documented
the fact that the
shielding
which
had
been installed
was effective in reducing radiation
levels to within acceptable
levels.
Unit 2
was
completed after
issuance
of the Bulletin and consequently,
shielding
was installed
during construction.
The licensee
stated
that
radiation
surveys
were performed in the fuel transfer tube area
and that,
although
the
surveys
indicated
no major problems,
additional
shielding
will be
added
to reduce
radiation levels at the
seal
between
the
Fuel
Handling Building
and
the
Reactor
Containment
Building to less
than
5 mr/hr.
Shielding will also
be installed to reduce the contact
dose rate
at the existing fuel transfer
tube shield structure in the annulus
region
to less
than
5 mr/hr.
After the additional
shielding is in place,
scheduled for installation in the Fall
1987, the fuel transfer
tube area
will again
be
surveyed.
These
additional
surveys will be reviewed
and
evaluated
in a future inspection
(50-335, 389/87-04-05).
ll.
Followup on
IE Information Notices
(92717)
The following IE Information Notices
were reviewed to ensure
receipt
and
review by appropriate
licensee
management.
86-20,
Low-Level Radioactive
Waste Scaling Factors,
86-22,
Underresponse
of Radiation
Survey Instrument to High Radiation
Fields
86-23,
Excessive
Skin Exposures
Due to Contamination
With Hot Particles
86-24, Respirator
Users Notice:
Increased
Inspection
Frequency for
Certain Self-Contained
Breathing Apparatus Air Cylinders
86-41, Evaluation of guestionable
Exposure
Readings of Licensee
Personnel
Dosimeters
86-42,
Improper Maintenance of Radiation Monitoring Systems
86-43,
Problems with Silver Zeolite Sampling of Airborne Radioiodine
86-44, Failure to Follow Procedures
When Working in High Radiation Areas
86-46,
Improper Cleaning
and Decontamination of Respiratory Protection
Equipment
13
86-55,
Delayed Access to Safety-Related
Areas
and Equipment During Plant
Emergencies86-103, Respirator
Coupling Nut Assembly Failures86-107,
Entry Into
PWR Cavity with Retractable
Incore Detector Thimbles
Withdrawn
87-03, Segregation
of Hazardous
and Low-Level Radioactive
Wastes
4
12.
Allegation Followup (99014)
Allegation
(RI I 85A0184)
During September
1985,
the alleger
was
employed
by Catalytic, Inc.
as
a
sheet
metal
worker at the St.
Lucie plant.
On September
26,
1985, at
about 7:00 a.m.
he
and five other workers were installing fire dampers
on
the -5 foot elevation of Unit
1 inside the
RCA under
RWP 85-267,
issued
on
July 15,
1985.
When the individual went to the control point to frisk out
of the controlled
area,
he
set off the frisker.
Health
Physics
(HP)
personnel
surveyed
him and
found
an unspecified
level of contamination
over his entire
body.
He was then told by
HP to sit down and wait to see
if the
contamination
decayed
away
because it was
probably
due
to
radioactive -noble gas
on his person.
The individual was subsequently
told
by his foreman
to go back into the
RCA an'd continue work.
At that point,
he told his foreman that
he would not go back into the area until someone
explained what had caused
the contamination
problem.
His foreman told him
that there
was
no other work to be performed
and that, if he would not go
back into the area,
he should
go home, which he did.
The next day he was
notified that
he
had been fired.
As
a result of this incident, the individual was concerned
that there
was
a noble
gas
problem in the area that was not properly controlled
and that
the licensee
had not filled out
a skin contamination or incident report.
Discussion
Through
records
review
the
inspector
found
the following additional
information:
(a)
The individual
had attended
a general
employee training
(GET) class
which
was
required for all
persons
working at the plant.
This
training was developed
to explain the various hazards of working at
a
nuclear
power plant including the subject of noble gas.
(b)
The licensee is required to post
an area
as
an airborne radioactivity
area
when it is
found that
the airborne activity is equal
to or
greater
than
25 percent
(X) of the
maximum permissible
concentration
'(MPC) listed in 10 CFR 20, Appendix B, Table 1, Column 1.
14
(c)
The licensee's
procedure,
HP-101, Identification
and
Reporting of
Radiation
Incidents,
Revision 2,
November 8,
1983,
required that
an
incident report be completed for personnel
contamination
in excess
of
100,000 disintegrations
per minute per one hundred
square
centimeters
(dpm/100 cm')
on skin or personal
clothing (5,000 dpm/100
cm~ on skin
or clothing
requires
documentation
on
a
personnel
skin/clothing
contamination
report).
Although not stated
in the procedure,
the
licensee
indicated
that this
does
not
apply
to
contamination
attributable
to noble
gas
because it decays
rapidly
and
dose
is
tracked
only
when concentration
of noble
gases
are in excess
of
5 mr/hr beta skin dose.
After discussions
with licensee
management
and
further
records
review,
the
inspector
determined
that
the airborne radioactivity
levels
on September
26,
1985,
were not in excess
of 25% of the
for noble
gas
but were
9.02%
MPC in the
-5 foot elevation
area.
Particulate
airborne
radioactivity levels
were
less
than
minimum
detectable
activity
(MDA).
Also, according
to the Health
Physics
Sign-in Sheet
for that date,
the alleger
was in the area
from 7:20
until 10:00 a.m. while other individuals,
who signed in at about the
same
time and were working in the
same
general
area,
remained there
until approximately
11:30 a.m.
Other personnel
in the area
had
been
contaminated,
as
had the alleger,
but
no contamination
was detected
on
anyone
upon exiting the
RCA for break/lunch.
The alleger
was
given
a whole body'ount
upon termination but no activity was detected.
Finding
The allegation
was partially substantiated
in that there
was
a noble
gas
problem
on September
26,
1985.
However,
the levels
were
such
that the area
was not required to be posted
as
an airborne area.
The
radiological
data
also indicated that the individual
was apparently"
contaminated
with noble
gas
which subsequently
decayed off.
Because
no contamination
was
detected
after the
noble
gas
had apparently
decayed
away,
the
licensee
was
not
required
to fill out
a
contamination
report.
No regulatory requirements
were violated
and
no deviations
were identified.
13.
Facility Statistics
Solid
Waste'uring
1986,
the
licensee
made
27
shipments
of radioactive
waste
consisting of 16,225 cubic feet of waste containing
2134.701
curies
of radioactivity.
This year to date,
6 shipments
had
been
made
consisting
of 1,934 cubic feet of waste
containing
a total of
496.321 curies of radioactivity.
15
b.
Contaminated
Area
C.
The licensee
began
tracking
square
footage of contaminated
area of
the plant. on February
1,
1986.
At that time 46,565
square feet or
approximately
35K was contaminated.
As of February 28, 1987, 33,763
square
feet or 25.3X remained
under contamination control.
Neither
reactor building was included in this inventory.
Personnel
Contamination
During 1986,
a total of 259 skin
and clothing contaminations
were
reported.
To date,
during 1987,
189 skin and clothing contamination
events
had been
documented.